![[logo] US EPA](http://www.epa.gov/epafiles/images/logo_epaseal.gif){kind=logo}
Biweekly Notice Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
Note: EPA no longer updates this information, but it may be useful as a reference or resource.
[Federal Register: December 12, 2001 (Volume 66, Number 239)]
[Notices]
[Page 64284-64312]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr12de01-145]
=======================================================================
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Biweekly Notice Applications and Amendments to Facility Operating
Licenses Involving No Significant Hazards Considerations
Note: The publication date for this notice will change from
every other Wednesday to every other Tuesday, effective January 8,
2002. The notice will contain the same information and will continue
to be published biweekly.
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the
[[Page 64285]]
pendency before the Commission of a request for a hearing from any
person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from November 19, 2001 through November 30, 2001.
The last biweekly notice was published on November 28, 2001 (66 FR
59498).
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, located at One White Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By January 11, 2002, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714, which is
available at the NRC's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible electronically from the
Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://
www.nrc.gov/NRC/ADAMS/index.html. If a request for a hearing or
petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
[[Page 64286]]
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Branch, or may be delivered to the Commission's Public
Document Room, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852, by the above date. A copy of
the petition should also be sent to the Office of the General Counsel,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to
the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Assess and Management Systems (ADAMS) Public Electronic
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/
NRC/ADAMS/index.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC Public Document room (PDR) Reference staff at 1-800-397-4209, 304-
415-4737 or by email to pdr@nrc.gov.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: July 5, 2001.
Description of amendment request: The proposed amendment would
relax Technical Specification (TS) operability requirements for primary
containment systems, secondary containment systems, and the standby gas
treatment system during the movement of irradiated fuel and during core
alterations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The equipment affected by the proposed changes are mitigative in
nature, and relied upon after an accident has been initiated.
Application of the Alternative Source Term (AST) does not involve a
change to the plant design. While the operation of the primary and
secondary containment systems do change as a result of these
proposed changes, these systems are not accident initiators.
Application of the AST does not initiate a design basis accident.
Similarly, application of the AST does not affect the design or
operation for any equipment or systems involved in the mitigation of
accidents. The proposed changes to the Technical Specifications
(TS), while they revise certain performance requirements, do not
involve any physical modifications to the plant. As a result, the
proposed changes do not affect any of the parameters or conditions
that could contribute to the initiation of any accidents. As such,
removal of operability requirements during the specified conditions
will not significantly increase the probability of occurrence for an
accident previously analyzed.
The AST changes do not affect the design and operation of the
facility. Rather, once the accident has been postulated the new
source term is an input to the evaluation of the consequences. The
implementation of the AST has been evaluated in revisions to the
analyses of the worst case Fuel Handling Accident (FHA) at Clinton
Power Station (CPS). Based on the results of the analyses, it has
been demonstrated that, with the proposed changes, the dose
consequences of the worst case FHA remain a small fraction of the
regulatory guidance provided by the NRC for the AST in RG
[regulatory guide]
1.183, ``Alternative Radiological Source Terms
for Evaluating Design Basis Accidents at Nuclear Power Reactors,''
dated July 2000. Since the primary containment systems, secondary
containment systems and the Standby Gas Treatment (SGT) are not
assumed to be operable in the FHA, the consequences of eliminating
the requirements that these systems be operable during the handling
of irradiated fuel in both primary and secondary containment or
during core alterations will not increase significantly.
In summary, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No new equipment is introduced, and no installed equipment is
operated in a new or different manner. There is no change to the
predicted accident response of any plant structure, system or
component. The proposed change in availability of mitigative
equipment has been evaluated in accordance with the guidance in RG
1.183 and does not produce different or more limiting accident
progression or results. As such, no new accident modes or equipment
failure modes are created by these proposed changes.
Therefore, these proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed changes involve a selective application of the AST
for the FHA consistent with the guidance provided in RG 1.183. The
existing analyses demonstrated that the dose consequences associated
with the FHA were within the applicable NRC specified limits. For
offsite dose, the margin to safety for the FHA using the 10 CFR 100,
``Reactor Site Criteria,'' limits was maintained by the existing
analysis. For the Control Room dose, the margin of safety using the
10 CFR 50, ``Domestic Licensing of Production and Utilization
Facilities,'' Appendix A, ``General Design Criteria for Nuclear
Power Plants,'' General Design Criteria 19, ``Control room,'' dose
limits was conservatively maintained by the existing analyses. The
results of the FHA analysis revised in support of this submittal
however, are subject to revised acceptance criteria. The revised
dose consequences of the limiting design basis FHA are within the
acceptance criteria found in RG 1.183 and 10 CFR 50.67, ``Domestic
Licensing of Production and Utilization Facilities, Accident Source
Term.'' The proposed changes ensure that the doses at the exclusion
area boundary (EAB), low population zone (LPZ), and control room
remain a small fraction of the new regulatory limits in RG 1.183 and
10 CFR 50.67.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Robert Helfrich, Mid-West Regional Operating
Group, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville,
IL 60555.
NRC Section Chief: Anthony J. Mendiola.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: October 31, 2001.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications (and, as applicable,
other elements of the licensing bases) to maintain a Post Accident
Sampling System (PASS). Licensees were generally required to implement
PASS upgrades as described in NUREG-0737,
[[Page 64287]]
``Clarification of TMI [Three Mile Island]
Action Plan Requirements,''
and Regulatory Guide 1.97, ``Instrumentation for Light-Water-Cooled
Nuclear Power Plants to Assess Plant and Environs Conditions During and
Following an Accident.'' Implementation of these upgrades was an
outcome of the lessons learned from the accident that occurred at TMI,
Unit 2. Requirements related to PASS were imposed by Order for many
facilities and were added to or included in the technical
specifications (TS) for nuclear power reactors currently licensed to
operate. Lessons learned and improvements implemented over the last 20
years have shown that the information obtained from PASS can be readily
obtained through other means or is of little use in the assessment and
mitigation of accident conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated October 31, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Richard P. Correia.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: October 31, 2001.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications (and, as applicable,
other elements of the licensing bases) to maintain a Post Accident
Sampling System (PASS). Licensees were generally required to implement
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three
Mile Island]
Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2. Requirements related to
PASS were imposed by Order for many facilities and were added to or
included in the technical specifications (TS) for nuclear power
reactors currently licensed to operate. Lessons learned and
improvements implemented over the last 20 years have shown that the
information obtained from PASS can be readily obtained through other
means or is of little use in the assessment and mitigation of accident
conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR
[[Page 64288]]
49271) on possible amendments to eliminate PASS, including a model
safety evaluation and model no significant hazards consideration (NSHC)
determination, using the consolidated line item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated October 31, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Richard P. Correia.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: May 24, 2001.
Description of amendment request: The proposed amendment would
delete License Condition 2.C.(11), which is no longer applicable to the
facility. License Condition 2.C.(11) requires inspection of the low-
pressure turbine discs during the second refueling outage, including
volumetric examination of the disc base using ultrasonic techniques,
and specifies that the frequency of subsequent inspections shall be in
accordance with the turbine manufacturer's recommendations. The
amendment request states that the license condition is no longer
applicable for the following reasons: (1) the initial inspection was
completed during the second refueling outage as required; and (2)
during fifth refueling outage, the low-pressure turbine rotors were
replaced with monoblock designed rotors that do not utilize shrunk-on
discs, and therefore the subsequent inspections specified in License
Condition 2.C.(11) for shrunk-on discs would be meaningless with the
new rotor design. The licensee's inspection and maintenance program for
the new low-pressure turbine is based on the current turbine
manufacturer's recommendations for the monoblock design.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment removes Fermi 2 Operating License
Condition 2.C.(11) which details the inspection frequency of the
low-pressure (LP) turbine discs. The inspection frequency was
recommended because the original turbine rotor design involved a
shrunk-on disc configuration. The inspection attributes applied
specifically to this disc design and were intended to enhance design
reliability. In 1996, however, the LP turbine steam path consisting
of rotors, buckets (blades), diaphragms and steam flow guides, all
manufactured by English Electric Co., were replaced with General
Electric (GE) components. In particular, the GE design does not
utilize shrunk-on discs; it includes rotors of monoblock
construction, thus negating the applicability of License Condition
2.C.(11). There are no relevant aspects of the
[[Page 64289]]
previously recommended inspections that apply to the new monoblock
construction.
Section 3.5.1.2.1 of the Fermi 2 UFSAR [Updated Final Safety
Analysis Report]
addresses the potential for missiles generated from
rotating equipment including those generated from a low-pressure
turbine rotor segment. Section 10.2.3 of the UFSAR states that
following the low-pressure turbine rotor replacement during RFO05,
``there will no longer be a design basis turbine missile at Fermi
2.'' Section 3.5.1.2.2 further states, ``The new low-pressure rotors
are of monoblock construction. The monoblock rotors have higher
speed capability than the maximum attainable speed of the turbine
generator units. Per General Electric, the supplier of the new
rotors, the probability of missiles being generated is well below 10
to the -8 power.'' There are no other postulated accidents that were
directly attributable to the English Electric Company shrunk-on disc
design; therefore, the removal of License Condition 2.C.(11) does
not increase the probability of occurrence or the consequences of
any accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change removes License Condition 2.C.(11) because
it is no longer applicable to the design of the low-pressure turbine
currently installed at the facility. Therefore, removal of the
license condition affects neither the design nor the operation of
the plant. It cannot create a new failure mode, nor can its removal
create the possibility of a new or different kind of accident than
any accident previously evaluated.
3. The change does not involve a significant reduction in the
margin of safety.
License Condition 2.C.(11) is not applicable to the facility
because the low-pressure turbine rotor was replaced with a design
which does not include shrunk-on turbine discs. This rotor
replacement eliminated the potential for a design basis accident
resulting from the turbine missiles at Fermi 2, which was the
accident scenario that the inspections referenced in License
Condition 2.C.(11) were intended to prevent. Since the license
condition no longer applies to the current facility design, and the
potential design basis accident associated with the license
condition no longer exists, the removal of the license condition
will not reduce any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Acting Section Chief: William D. Reckley, Acting.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: November 11, 2001.
Description of amendment request: A change is proposed to Technical
Specification 3.0.3 to allow a longer period of time to perform a
missed surveillance. The time is extended from the current limit of ``
* * * up to 24 hours or up to the limit of the specified Frequency,
whichever is less'' to ``* * *up to 24 hours or up to the limit of the
specified Frequency, whichever is greater.'' In addition, the following
requirement would be added to the specification: ``A risk evaluation
shall be performed for any Surveillance delayed greater than 24 hours
and the risk impact shall be managed.''
The Nuclear Regulatory Commission (NRC) staff issued a notice of
opportunity for comment in the Federal Register on June 14, 2001 (66 FR
32400), on possible amendments concerning missed surveillances,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on September 28, 2001 (66 FR
49714). The licensee affirmed the applicability of the following NSHC
determination in its application dated November 11, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO [Limiting Condition for
Operation]
is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on the margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
[[Page 64290]]
NRC Section Chief: William D. Reckley, Acting.
Dominion Nuclear Connecticut, Inc., et al., Docket Nos. 50-245, 50-336,
and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New
London County, Connecticut
Date of amendment request: November 8, 2001.
Description of amendment request: The proposed amendments would
incorporate administrative and editorial changes into the Millstone
Unit No. 1 Permanently Defueled Technical Specifications (PDTS) and
into the Millstone Unit Nos. 2 and 3 Technical Specifications (TSs).
Specifically, the proposed changes would: (1) Relocate redundant design
features information already included in other licensing basis (LB)
documents (e.g., the Final Safety Analysis Report (FSAR)), from Section
5.0, ``Design Features,'' of the Unit Nos. 2 and 3 TS, to other LB
documents, consistent with the improved Standard Technical
Specifications (STSs) for the respective unit design; (2) revise TS
5.6.2, ``Technical Specifications Bases Control Program,'' in the Unit
No. 1 PDTS to incorporate the 10 CFR 50.59 rule change; (3) add a new
TS (TS 6.22 for Unit No. 2 and TS 6.17 for Unit No. 3), to incorporate
a TS bases control program within the Unit Nos. 2 and 3 TS; (4) add a
new TS (TS 6.18, ``Component Cyclic or Transient Limits''), to the Unit
No. 3 TS to define the program for tracking cyclic (or transient)
limits. These limits are proposed to be relocated from where they are
listed in TS 5.7, ``Component Cyclic or Transient Limit,'' in the Unit
No. 3 TS, to the FSAR; (5) revise the Unit No. 1 PDTS and the Unit Nos.
2 and 3 TS related to Radiological Environmental Monitoring Program
(REMP) procedure processing to: (a) remove reference to an organization
affiliated with Northeast Utilities (NU), the Production Operations
Services Laboratory, which is no longer applicable following the change
in ownership from NU to Dominion Nuclear Connecticut (DNC); (b) replace
the reference to the Radiological Assessment Branch (a Millstone DNC
organization) with the ``organization responsible for the REMP'' for
review/approval of changes to the REMP to avoid future TS changes due
to a change in organizational titles; (c) correct an inconsistency
within the Unit No. 1 PDTS which implies that REMP procedures are
processed under the general procedure processing specification (i.e.,
TS 5.5.1), in addition to the specific specifications for processing
REMP procedure changes (i.e., Specifications 5.5.6 and 5.5.7); and (6)
correct miscellaneous editorial issues and achieve consistency between
the TSs for each unit. These changes include: (a) Changes to and
corrections in titles; (b) correct references to the quality assurance
program, and (c) change titles to utilize the term radiation protection
rather than health physics.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes related to Section 5, ``Design Features,''
of the Unit Nos. 2 or 3 TS either relocates or deletes certain
details from the Technical Specifications and relocates them to the
respective unit's updated FSAR or other plant controlled documents.
The FSAR and other plant controlled documents will be maintained in
accordance with 10 CFR 50.59. The proposed changes to Section 6,
``Administrative Controls,'' adds new administrative specifications
consistent with the guidance of the improved STS, corrects
inconsistencies, or represents changes in nomenclature, and will
correct editorial issues and achieve consistency within the
individual TS and between individual TS. The changes are purely
administrative or editorial and do not alter any regulatory
requirements or have an impact on the acceptance criteria for any
design basis accident described in the respective Unit Nos. 2 or 3
FSAR or the Unit No. 1 Defueled Safety Analysis Report (DSAR).
These changes have no impact on plant equipment operation. Since
the changes are solely an administrative or editorial change to the
TS, they cannot affect the likelihood or consequences of accidents.
Therefore, these changes will not increase the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes have no impact on plant operation. Since
the proposed changes are solely an administrative or editorial
change to the TS, they do not affect plant operation in any way. The
proposed changes do not involve a physical alteration of the plant
or change the plant configuration (no new or different type of
equipment will be installed). The proposed changes do not require
any new or unusual operator actions. The changes do not alter the
way any structure, system, or component functions and do not alter
the manner in which the plant is operated. The changes do not
introduce any new failure modes. Therefore, the proposed changes
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Since the proposed changes are solely administrative or
editorial changes to the TS, they do not affect plant operation in
any way. The proposed changes to the respective unit's technical
specifications will standardize terminology, remove extraneous
information and make minor format changes that will not result in
any technical changes to current requirements.
The proposed changes do not impact any acceptance criteria for
the design basis accidents described in the respective Unit Nos. 2
or 3 FSAR or the Unit No. 1 DSAR and do not impact the consequences
of accidents previously evaluated. Therefore, the proposed changes
will not result in a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Section Chief: James W. Clifford.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: May 25, 2001.
Description of amendment request: The amendments would revise
Technical Specifications (TS) Definitions for ENGINEERED SAFETY FEATURE
(ESF) RESPONSE TIME and REACTOR TRIP SYSTEM (RTS) RESPONSE TIME to
provide for verification of response time for selected components
provided that the components and the methodology for verification have
been previously reviewed and approved by the Nuclear Regulatory
Commission. The associated Bases will also be revised. The licensee has
referenced previously approved WCAP-13632-P-A, Revision 2,
``Elimination of Pressure Sensor Response Time Testing Requirements,''
and WCAP-14036-P-A Revision 1, ``Elimination of Periodic Protection
Channel Response Time Tests'' as the justifications for proposing these
changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 64291]]
Conformance of the proposed amendments to the standards for a
determination of no significant hazards as defined in 10 CFR 50.92
is shown in the following:
(1) The proposed license amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This change to the TS does not result in a condition where the
design, material, and construction standards that were applicable
prior to the change are altered. The same RTS and ESFAS
instrumentation is being used; the time response allocations/
modeling assumptions in the UFSAR Chapter 15 analyses are still the
same; only the method of verifying time response is changed. The
proposed change will not modify any system interface and could not
increase the likelihood of an accident since these events are
independent of this change. The proposed activity will not change,
degrade, or prevent actions or alter any assumptions previously made
in evaluating the radiological consequences of an accident described
in the UFSAR. Therefore, the proposed amendments do not result in
any increase in the probability or consequences of an accident
previously evaluated.
(2) The proposed license amendments do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
This change does not alter the performance of the reactor
protection system (RPS) or the engineered safety features actuation
system (ESFAS). All RPS and ESFAS channels will still have response
time verified by test before placing the channel in operational
service and after any maintenance that could affect response time.
Changing the method of periodically verifying instrument response
for certain RPS and ESFAS channels (assuring equipment operability)
from time response testing to calibration and channel checks will
not create any new accident initiators or scenarios. Periodic
surveillance of these instruments will detect significant
degradation in the channel characteristic. Implementation of the
proposed amendments does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) The proposed license amendments do not involve a significant
reduction in a margin of safety.
This change does not affect the total system response time
assumed in the safety analysis. The periodic system response time
verification method is modified to allow use of actual test data or
engineering data. The method of verification still provides
assurance that the total system response is within that defined in
the safety analysis, since calibration tests will detect any
degradation which might significantly affect channel response time.
Based on the above, it is concluded that the proposed license
amendment request does not result in a reduction in a margin with
respect to plant safety.
Based on the preceding analysis, it is concluded that
elimination of periodic [response time testing]
RTT is acceptable
and the proposed license amendments do not involve a significant
hazards consideration finding as defined in 10 CFR 50.92.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: Richard J. Laufer, Acting.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: August 6, 2001.
Description of amendment request: The amendments would revise
Technical Specifications (TS) 3.3.2 for engineered safety feature
actuation system instrumentation, TS 3.3.6 for containment purge and
exhaust isolation instrumentation. The amendments would also revise the
appropriate bases, and the bases for Containment Isolation Valves (TS
3.6.3). Specifically, the proposed amendments would modify the TS
requirements so that they exclude the Containment Purge Ventilation
System and the Hydrogen Purge System, thereby applying the requirements
to only the Containment Air Release and Addition System. At Catawba,
the containment isolation valves for the Containment Purge Ventilation
System and the Hydrogen Purge System are sealed closed in the modes of
applicability (Modes 1, 2, 3, and 4) according to TS requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following discussion is a summary of the evaluation of the
changes contained in this proposed amendment against the 10 CFR
50.92(c) requirements to demonstrate that all three standards are
satisfied. A no significant hazards consideration is indicated if
operation of the facility in accordance with the proposed amendment
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, or
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3. Involve a significant reduction in a margin of safety.
First Standard
Implementation of this amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Neither the Containment Purge Ventilation
System, the Hydrogen Purge System, nor the Containment Air Release
and Addition System is capable of by itself initiating any accident.
The safety related portions of these systems, which are responsible
for maintaining containment isolation during accident conditions,
will continue to function as designed, and in accordance with all
applicable TS. The design and operation of the systems are not being
modified by this proposed amendment. Therefore, there will be no
impact on any accident probabilities or consequences.
Second Standard
Implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new accident causal mechanisms are created
as a result of NRC approval of this amendment request. No changes
are being made to the plant which will introduce any new accident
causal mechanisms. This amendment request does not impact any plant
systems that are accident initiators and does not impact any safety
analyses.
Third Standard
Implementation of this amendment would not involve a significant
reduction in a margin of safety. Margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design functions during and following an accident situation.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. The performance of these fission
product barriers will not be impacted by implementation of this
proposed amendment. It has already been shown that the performance
of all containment isolation functions in response to accident
conditions will not be impacted by this proposed amendment. There is
no risk significance to this proposed amendment, as no reduction in
system or component availability will be incurred. No safety margins
will be impacted.
Based upon the preceding discussion, Duke Energy has concluded
that the proposed amendment does not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: Richard J. Laufer, Acting.
[[Page 64292]]
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: September 24, 2001.
Description of amendment request: The amendment request proposes to
extend the allowed outage time for a Division I or Division II
Emergency Diesel Generator (EDG) from 72 hours to 14 days. The proposed
changes are intended to provide flexibility in scheduling EDG
maintenance activities, reduce refueling outage duration, and improve
EDG availability during plant shutdowns.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
The proposed Technical Specification (TS) changes do not affect
the design, operational characteristics, function, or reliability of
the EDGs. The EDGs are not the initiators of previously evaluated
accidents. The EDGs are designed to mitigate the consequences of
previously evaluated accidents including a loss of offsite power.
Extending the allowed outage time (AOT) for a single EDG would not
significantly affect the previously evaluated accidents since the
remaining EDGs supporting the redundant ESF [Engineered Safety
Feature]
systems would continue to perform the accident mitigating
functions as designed.
The duration of a TS AOT is determined considering that there is
a minimal possibility that an accident will occur while a component
is removed from service. A risk-informed assessment was performed
which concluded that the increase in plant risk is small and
consistent with the USNRC [United States Nuclear Regulatory
Commission (NRC)]
``Safety Goals for the Operations of Nuclear Power
Plants; Policy Statement,'' Federal Register, Vol. 51, p.30028 (51
FR 30028), August 4, 1986, as further described by NRC Regulatory
Guide 1.177.
The current TS requirements establish controls to ensure that
redundant systems relying on the remaining EDGs are Operable. In
addition to these requirements, administrative controls will be
established to provide assurance that the AOT extension is not
applied during adverse weather conditions that could potentially
affect offsite power availability.
Both the RBS [River Bend Station, Unit 1]
risk-based analysis
and the deterministic evaluation support the increased AOT.
Therefore, this change does not involve a significant increase in
the probability or consequences of any accident previously
evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
The proposed TS changes do not involve a change in the design,
configuration, or method of operation of the plant that could create
the possibility of a new or different kind of accident. The proposed
change extends the AOT currently allowed by the TS.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
The proposed extended AOT is not in conflict with any of the
approved codes and standards applicable to the onsite AC
[Alternating Current]
power sources. The proposed changes do deviate
from the recommendations of Regulatory Guide (RG) 1.93. An extension
of the 72 hour AOT recommended in the RG to 14 days is demonstrated
herein to be acceptable and has been approved for several other
licensees. Assuming there are no additional failures of redundant
equipment during the time that the EDG is removed from service, the
intended safety functions would still be met.
The proposed AOT change does not affect any of the assumptions
or inputs to the safety analyses of the FSAR [Final Safety
Assessment Report]
and does not erode the decrease in severe
accident risk achieved with the issuance of the Station Blackout
(SBO) Rule, 10 CFR 50.63 ``Loss of All Alternating Current Power''.
RBS is classified as a four-hour coping plant with 0.95 EDG
reliability (see U[pdated]
FSAR Appendix 15C). The assumptions used
in the SBO [Station Blackout]
analysis regarding reliability of the
EDGs are unaffected by the proposed TS changes since preventive
maintenance and testing will continue to be performed to maintain
reliability assumptions.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit Number 3, Westchester County, New York
Date of amendment request: October 23, 2001.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.5.10, ``Ventilation Filter
Testing Program,'' to adopt the requirements of the American Society
for Testing and Materials Standard (ASTM) D3803-1989, ``Standard Test
Method for Nuclear-Grade Activated Carbon.'' The proposed TS revisions
are in response to Nuclear Regulatory Commission (NRC) Generic Letter
99-02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal.'' The
proposed amendment would also revise the differential pressure criteria
for the test of the filter system for the Control Room Ventilation
System.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed license amendment adopts the new test method and
acceptance criteria of ASTM D3803-1989, with the exceptions
identified, for activated charcoal filters and changes the allowable
pressure differential for Control Room ventilation. The changes
require laboratory performance testing of adsorber carbon that
yields a more accurate result than the testing currently required by
the TS and requires a more stringent limit on the Control Room
ventilation pressure differential. The proposed change to delete
non-conservative TS requirements for testing of adsorber carbon and
limiting the Control Room ventilation differential pressure are not
plant accident initiators as described in the Final Safety Analysis
Report (FSAR). The proposed amendment does not change the function
of any structure, system or component (SSC). The function of the
ventilation systems is filtration of radiological releases during
postulated accidents. The proposed changes will provide greater
assurance that this function is provided. The revised TS
requirements are for laboratory tests and pressure differential
measurements that are currently in place and change only the TS
testing requirements. They will not result in any changes to the
efficiency assumed in accident analysis. The changes do not alter,
degrade or prevent actions described or assumed in an accident
described in the FSAR. Therefore, the proposed amendment does not
change the possibility of an accident previously evaluated or
significantly increase the consequences of an accident previously
evaluated.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
The proposed license amendment adopts the new test method and
acceptance criteria of ASTM D3803-1989, with the exceptions
[[Page 64293]]
identified, for activated charcoal filters and changes the allowable
pressure differential for Control Room ventilation. The change does
not involve any modifications to the plant, will not require changes
to how the plant is operated nor will it affect the operation of the
plant. The changes require laboratory performance testing of
adsorber carbon that yields a more accurate result than the testing
currently required by the TS and requires a more stringent limit on
the Control Room ventilation pressure differential. The proposed
changes to delete non-conservative TS requirements for testing of
adsorber carbon and limiting the Control Room ventilation
differential pressure are not plant accident initiators as described
in the Final Safety Analysis Report (FSAR). The proposed amendment
does not change the function of any structure, system or component
(SSC). The function of the ventilation systems is filtration of
radiological releases during postulated accidents. The proposed
changes will provide greater assurance that this function is
provided. The revised TS requirements are for laboratory tests and
pressure differential measurements that are currently in place and
change only the TS testing requirements. They will not result in any
changes to the efficiency assumed in accident analysis. The changes
do not alter, degrade or prevent actions described or assumed in an
accident described in the FSAR. Therefore, the proposed amendment
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Does the proposed license amendment involve a significant
reduction in a margin of safety?
The proposed license amendment adopts the new test method and
acceptance criteria of ASTM D3803-1989, with the exceptions
identified, for activated charcoal filters and changes the allowable
pressure differential for Control Room ventilation. The proposed
license amendment does not reduce the margin of safety but enhances
by requiring more accurate testing and a more conservative pressure
differential. The proposed test change will require the use of a
current and improved ASTM standard to ensure that the carbon ability
to adsorb radioactive material will remain at or above the
capability credited in our accident analysis. The proposed
differential pressure limit will assure that the system provides
sufficient flow though the charcoal to meet accident analyses.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: L. Raghavan (Acting).
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: October 23, 2001.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications (TSs) (and, as
applicable, other elements of the licensing bases) to maintain a Post
Accident Sampling System (PASS). Licensees were generally required to
implement PASS upgrades as described in NUREG-0737, ``Clarification of
TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades was an outcome of the
lessons learned from the accident that occurred at TMI, Unit 2.
Requirements related to PASS were imposed by Order for many facilities
and were added to or included in the TSs for nuclear power reactors
currently licensed to operate. Lessons learned and improvements
implemented over the last 20 years have shown that the information
obtained from PASS can be readily obtained through other means or is of
little use in the assessment and mitigation of accident conditions.
The Nuclear Regulatory Commission (NRC) staff issued a notice of
opportunity for comment in the Federal Register on August 11, 2000 (65
FR 49271) on possible amendments to eliminate PASS, including a model
safety evaluation and model no significant hazards consideration (NSHC)
determination, using the consolidated line item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated October 23, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or
[[Page 64294]]
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in [a]
Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in [a]
margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: L. Raghavan (Acting).
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: September 28, 2001.
Description of amendment request: The licensee proposes to revise a
single Anticipated Transient Without Scram (ATWS) Recirculation Pump
Trip Reactor Pressure High setpoint to replace the current conditional
setpoints which are based upon the number of Safety Relief Valves out
of service.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated because: a change
in the ATWS high RPV [reactor pressure vessel]
pressure RWR [ATWS
Reactor Pressure High Recirculation Pump]
pump trip setpoint does
not affect initiation of any accident. Operation in accordance with
the revised setpoint ensures the consequences of previously analyzed
accidents are not changed.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated because: RPV pressure
following an ATWS with PRFO [Pressure Regulating Valve Open]
event
(worst case transient for RPV pressurization) remains within
acceptable limits with the revised setpoint. Therefore, changing the
setpoint will not lead to a new or different kind of accident.
3. Involve a significant reduction in a margin of safety
because: the analyses performed to determine the revised ATWS high
pressure RWR pump trip setpoint assure maintenance of the same
margin of safety as presently exists for limiting RPV pressure
following an ATWS with PRFO (limiting transient). The current
analyses actually shows an improved margin over the results of the
previous analyses (References 2 and 3), which were performed using
an earlier computer code.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: L. Raghavan (Acting).
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: October 30, 2001.
Description of amendment request: The license amendment request
proposes changes to Arkansas Nuclear One, Unit 2 (ANO-2) Technical
Specification (TS) 3.4.9, ``Pressure/Temperature Limits,'' and TS
3.4.12, ``Low Temperature Overpressure Protection (LTOP) System.'' The
primary changes are to update the existing pressure/temperature (P/T)
limits from 21 to 32 effective full power years (EFPYs) and to include
additional restrictions in the LTOP TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--Does Not Involve a Significant Increase in the Probability
or Consequences of an Accident Previously Evaluated.
The probability of occurrence of an accident previously
evaluated for ANO-2 is not altered by the proposed amendment to the
technical specifications (TSs). The accidents remain the same as
currently analyzed in the ANO-2 Safety Analysis Report (SAR) as a
result of changes to the P/T limits as well as those for LTOP. The
new P/T and LTOP limits were based on NRC [Nuclear Regulatory
Commission]
accepted methodologies along with ASME [American Society
of Mechanical Engineers]
Code [Boiler and Pressure Vessel Code]
alternatives. The proposed changes do not impact the integrity of
the reactor coolant pressure boundary (RCPB) (i.e. there is no
change to the operating pressure, materials, loadings, etc.) as a
result of this change. In addition, there is no increase in the
potential for the occurrence of a loss of coolant accident. The
probability of any design basis accident is not affected by this
change, nor are the consequences of any design basis accident (DBA)
affected by this proposed change. The proposed P/T limit curves and
the LTOP limits are not considered to be an initiator or contributor
to any accident currently evaluated in the ANO-2 SAR. These new
limits ensure the long term integrity of the RCPB.
Fracture toughness test data are obtained from material
specimens contained in capsules that are periodically withdrawn from
the reactor vessel. These data permit determination of the
conditions under which the vessel can be operated with adequate
safety margins against non-ductile fracture throughout its service
life. A new reactor vessel specimen capsule was withdrawn at the
most recent refueling outage and was analyzed to predict the
fracture toughness requirements using projected neutron fluence
calculations. For each analyzed transient and steady state
condition, the allowable pressure is determined as a function of
reactor coolant temperature considering postulated flaws in the
reactor vessel beltline, inlet nozzle, outlet nozzle, and closure
head.
The predicted radiation induced DRTNDT [shift
in reference temperature for nil-ductility transition]
was
calculated using the respective reactor vessel beltline materials
copper and nickel contents and the neutron fluence applicable to 32
EFPY including an estimated increase in flux due to a proposed power
uprate. The DRTNDT [reference temperature for
nil-ductility transition]
and, in turn, the operating limits for
ANO-2 were adjusted to account for the effects of irradiation on the
fracture toughness of the reactor vessel materials. Therefore, new
operating limits are established which are represented in the
revised operating curves for heatup/criticality, cooldown and
inservice hydrostatic testing contained in the technical
specifications.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2--Does Not Create the Possibility of a New or Different Kind
of Accident From Any Previously Evaluated
The proposed changes to the P/T and LTOP limits will not create
a new accident scenario. The requirements to have P/T and LTOP
protection are part of the licensing basis of ANO-2. The proposed
changes
[[Page 64295]]
reflect the change in vessel material properties acknowledged and
managed by regulation and the best data available in response to NRC
Generic Letter 92-01, Revision 1. The approach used meets NRC and
ASME regulations and guidelines. The calculational methodology for
fluence is based on an NRC approved Framatome ANP approach.
Therefore, the adjusted reference temperatures for fracture
toughness are consistent with that previously provided to the NRC.
The data analysis for the vessel specimen removed at 2R14
(approximately 15.7 EFPY of exposure) confirms that the vessel
materials are responding as predicted.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the Margin of
Safety
The existing P/T curves and LTOP limits in the technical
specifications are reaching their expiration period for the number
of years at effective full power operation. The revision of the P/T
limits and curves will ensure that ANO-2 continues to operate within
the operating margins allowed by 10 CFR 50.60 and the ASME Code. The
material properties used in the analysis are based on results
established through CE [Combustion Engineering]
material reports for
copper and nickel content. The application of ASME Code Case N-641
presents alternative procedures for calculating P/T and LTOP
temperatures and pressures in lieu of that established for ASME
Section XI, Appendix G-2215. This Code alternative allows certain
assumptions to be conservatively reduced. However, the procedures
allowed by Code Case N-641 still provide significant conservatism
and ensure an adequate margin of safety in the development of P/T
operating and pressure test limits to prevent non-ductile fractures.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of amendment request: November 19, 2001.
Description of amendment request: The proposed amendment would to
eliminate restrictions imposed by technical specification (TS) 3.0.4
for the Remote Shutdown Instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No.
Probability of Occurrence of an Accident Previously Evaluated
The Remote Shutdown Instrumentation system ensures that
sufficient capability is available to permit shutdown and
maintenance of Hot Standby of the plant from locations outside of
the control room. The proposed change allows Unit 1 to ascend in
mode without meeting the LCO [limiting condition for operation]
for
TS 3.3.3.5. The proposed change does not impact the ability to
comply with the allowed outage time (AOT) described in TS 3.3.3.5.
As such, the proposed change does not affect any accident initiators
or precursors, since the AOT for TS 3.3.3.5 will continue to be met.
The proposed change is also consistent with the Unit 2 TS.
Therefore, the probability of occurrence of an accident previously
evaluated is not significantly increased.
The format changes do not impact any accident initiators or
precursors. Thus, the probability of occurrence of an accident
previously evaluated is not significantly increased.
Consequences of an Accident Previously Evaluated
The proposed change to allow Unit 1 to ascend in mode without
meeting the LCO for TS 3.3.3.5, while continuing to meet the action
statement, will not significantly impact the Remote Shutdown
Instrumentation systems' capability of performing its design
function. The Remote Shutdown Instrumentation ensures that
sufficient capability is available to permit shutdown and
maintenance of Hot Standby of the plant from locations outside of
the control room. The proposed change does not impact the ability to
comply with AOT described in TS 3.3.3.5. The proposed change is also
consistent with the Unit 2 TS. Thus, there will be no increase in
offsite doses, and the consequences of an accident previously
analyzed are not increased.
The format changes do not impact the function of the Remote
Shutdown Instrumentation. Thus, there will be no increase in offsite
doses, and the consequences of an accident previously analyzed are
not significantly increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The Remote Shutdown Instrumentation system ensures that
sufficient capability is available to permit shutdown and
maintenance of Hot Standby of the plant from locations outside of
the control room. Allowing Unit 1 to ascend in mode without meeting
the LCO for TS 3.3.3.5, while continuing to meet the action
statement, does not change the function of the Remote Shutdown
Instrumentation system or create the possibility of a new or
different type of accident. The proposed change does not impact the
ability to comply with the AOT described in TS 3.3.3.5. The proposed
change is also consistent with the Unit 2 TS. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
The format changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not impact the Remote Shutdown
Instrumentation system's capability of performing its design
function, nor does the proposed change impact the operational
characteristics of the Remote Shutdown Instrumentation system. The
Remote Shutdown Instrumentation ensures that sufficient capability
is available to permit shutdown and maintenance of Hot Standby of
the plant from locations outside of the control room. Allowing Unit
1 to ascend in mode without meeting the LCO for TS 3.3.3.5, while
continuing to meet the action statement, does not impact CNP's
accident analysis. The proposed change is also consistent with the
Unit 2 TS. Therefore, the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: William D. Reckley, Acting.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: October 12, 2001.
Description of amendment requests: The proposed amendments would
delete requirements from the technical specifications (TSs) (and, as
applicable, other elements of the licensing bases) to maintain a Post
Accident Sampling System (PASS). Licensees were generally required to
implement PASS upgrades as described in NUREG-0737, ``Clarification of
TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades was
[[Page 64296]]
an outcome of the lessons learned from the accident that occurred at
TMI, Unit 2. Requirements related to PASS were imposed by Order for
many facilities and were added to or included in the TSs for nuclear
power reactors currently licensed to operate. Lessons learned and
improvements implemented over the last 20 years have shown that the
information obtained from PASS can be readily obtained through other
means or is of little use in the assessment and mitigation of accident
conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated October 12, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Therefore, the NRC staff proposes to determine that the amendment
requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: William D. Reckley, Acting.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: November 1, 2001.
Description of amendment requests: The proposed amendments would
revise technical specification (TS) surveillance requirements (SR)
4.8.2.3.2.c.2 and 4.8.2.5.2.c.2 and associated TS bases concerning the
safety-related batteries to make them more consistent with the
Westinghouse Standard TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
Probability of Occurrence of an Accident Previously Evaluated
The proposed change to SRs 4.8.2.3.2.c.2 and 4.8.2.5.2.c.2 to
add a requirement to remove visible corrosion and to delete the
requirement that the battery be free of corrosion does not affect
any accident initiators or precursors. The batteries perform a
mitigating function following a loss of AC power, and the presence
of corrosion will not adversely impact components whose failure
would initiate an accident. Thus, the probability of occurrence of
an accident previously evaluated is not significantly increased.
The proposed change to the TS 3/4.8 bases provides clarification
and does not affect any accident initiators or precursors. Thus, the
probability of occurrence of an accident previously evaluated is not
significantly increased.
The proposed change to SRs 4.8.2.3.2.c.3 and 4.8.2.5.2.c.3
increases the battery charger current required during surveillance
testing. The required value is within the capability of
[[Page 64297]]
the battery charger. Thus, the battery charger is not degraded by
this change, and the change does not affect any accident initiators
or precursors. Thus, the probability of occurrence of an accident
previously evaluated is not significantly increased.
The proposed changes to SR 4.8.2.3.2.d delete the requirement
that the battery terminal voltage be maintained greater than or
equal to 210 volts during the battery service test, and delete the
description of the composite load profile. The removal of the
requirement and the description from the SR do not affect any
accident initiators or precursors. Thus, the probability of
occurrence of an accident previously evaluated is not significantly
increased.
The deletion of Tables 4.8-2 and 4.8-3, the incorporation of the
words ``this page intentionally left blank,'' and the deletion of
the SR 4.8.2.3.2.d and SR 4.8.2.5.2.d references to the tables do
not impact battery operation as the tables summarize information
used as calculation inputs. These changes do not affect any accident
initiators or precursors. Thus, the probability of occurrence of an
accident previously evaluated is not significantly increased.
The proposed changes to SR 4.8.2.5.2.d to delete the requirement
that the battery terminal voltage be maintained greater than or
equal to 210 volts during the battery service test, and to add the
term ``design duty cycle'' does not affect any accident initiators
or precursors. Thus, the probability of occurrence of an accident
previously evaluated is not significantly increased.
The editorial change does not impact any accident initiators or
precursors. Thus, the probability of occurrence of an accident
previously evaluated is not significantly increased.
Consequences of an Accident Previously Evaluated
The batteries and their associated chargers provide power to
emergency equipment that is used in the mitigation of accidents. The
batteries provide power to this equipment following a loss of AC
power until the battery chargers are powered by the emergency diesel
generators.
The proposed change to SRs 4.8.2.3.2.c.2 and 4.8.2.5.2.c.2 to
add a requirement to remove visible corrosion and to delete the
requirement that the battery connections be free of corrosion does
not impact a battery's capability to power its safety-related loads
as the presence of corrosion at the terminal connections does not
indicate that the battery is unable to perform its function. Rather,
it is the impact of the corrosion on the connections that is of
concern. This concern will be addressed by performing a resistance
check to verify that battery performance is acceptable. Therefore,
this change does not result in an increase in offsite doses. Thus,
the consequences of an accident previously analyzed are not
increased.
The proposed change to the TS 3/4.8 bases provides clarification
and does not impact the battery's capability to power its safety-
related loads. Thus, the consequences of an accident previously
analyzed are not increased.
The proposed change to SRs 4.8.2.3.2.c.3 and 4.8.2.5.2.c.3 to
increase the required battery charger current ensures that the
battery charger has sufficient capacity to provide power to
emergency equipment while simultaneously recharging batteries that
were discharged following a loss of AC power. This ensures that
emergency equipment connected to the battery will continue to
operate as designed, and offsite doses will not be increased. Thus,
the consequences of an accident previously analyzed are not
increased.
The proposed changes to SR 4.8.2.3.2.d delete the requirement
that the battery terminal voltage be maintained greater than or
equal to 210 volts during the battery service test, and delete the
description of the composite load profile. However, the SR will
still require that the service test demonstrate that the battery
capacity is adequate to supply emergency loads. The voltage
requirements for the batteries are determined by battery-system
specific calculations, and the calculation results are incorporated
into the test procedures. This assures that the equipment connected
to the battery will continue to operate as designed, and offsite
doses will not be increased. Thus, the consequences of an accident
previously analyzed are not increased.
The deletion of Tables 4.8-2 and 4.8-3, the addition of the
words ``this page intentionally left blank,'' and the deletion of
the SR 4.8.2.3.2.d and SR 4.8.2.5.2.d references to the tables do
not impact battery operation as the tables summarize information
used as calculation inputs. The batteries are tested to a load
profile that is developed on the basis of the battery loads for a
loss of AC power, and the testing assures that the batteries are
capable of performing their safety function. Thus, these changes
will not impact battery capability, will not result in an increase
in offsite doses, and the consequences of an accident previously
analyzed are not increased.
The proposed changes to SR 4.8.2.5.2.d to delete the requirement
that the battery terminal voltage be maintained greater than or
equal to 210 volts during the battery service test, and to add the
term ``design duty cycle'' requires that the battery be tested in
accordance with a load profile developed on the basis of the battery
loads for a loss of AC power. The testing of the battery assures
that it is capable of performing its safety function. Thus, the
capability of the battery is not impacted, there will be no increase
in offsite doses, and the consequences of an accident previously
analyzed are not increased.
The editorial change does not impact battery capability. Thus,
there will be no increase in offsite doses, and the consequences of
an accident previously analyzed are not increased.
Therefore, the probability of occurrence or the consequences of
accidents previously evaluated are not increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The batteries perform a mitigating function by providing power
to emergency equipment following a loss of AC power.
The proposed change to SRs 4.8.2.3.2.c.2 and 4.8.2.5.2.c.2 adds
a requirement to remove visible corrosion and deletes the
requirement that the battery terminals be free of corrosion. The
presence of corrosion on the battery terminals does not introduce a
mechanism that would cause a plant transient, and I&M will ensure
that the corrosion does not impact the battery's function. Thus, the
possibility of a new or different kind of accident is not created.
The proposed change to the TS 3/4.8 bases provides clarification
and does not introduce a mechanism that would cause a plant
transient. Thus, the possibility of a new or different kind of
accident is not created.
The proposed change to SRs 4.8.2.3.2.c.3 and 4.8.2.5.2.c.3
increases the acceptance criterion for battery charger current to
reflect the present demand on the battery charger when it is
simultaneously supplying power to emergency equipment and charging a
discharged battery. The increase in the acceptance criterion is
within the capability of the battery charger, and no failure
mechanisms are introduced by this change. Thus, the change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
The proposed changes to SR 4.8.2.3.2.d to delete the requirement
that the battery terminal voltage be maintained greater than or
equal to 210 volts during a battery service test, and to delete the
load profile description do not directly impact any emergency
equipment as the SR continues to require that the battery service
test demonstrate that the battery is capable of supplying power to
connected equipment, and this change does not introduce any battery
failure mechanisms. Thus, the change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
The deletion of Tables 4.8-2 and 4.8-3, the incorporation of the
words ``this page intentionally left blank,'' and the deletion of
the SR 4.8.2.3.2.d and SR 4.8.2.5.2.d references to the tables do
not impact battery operation as the tables summarize information
used as calculation inputs. Thus, the changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
The proposed changes to SR 4.8.2.5.2.d to delete the requirement
that the battery terminal voltage be maintained greater than 210
volts during a battery service test, and to add the term ``design
duty cycle'' do not introduce any battery failure mechanisms as they
do not alter the battery's physical characteristics or the battery
testing requirements. Additionally, the term ``design duty cycle''
more accurately reflects the use of a simulated load for the battery
test. Thus, the change does not create the possibility of a new or
different kind of accident from any previously evaluated.
The editorial change does not create the possibility of a new or
different kind of accident from any previously evaluated.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes do not impact the functional requirements
of either the
[[Page 64298]]
batteries or the battery chargers, nor do the changes impact the
operational characteristics of the equipment that is connected to
the battery. The batteries will continue to be subjected to a system
test to verify that the battery capacity is adequate, and the
battery chargers will be tested to verify that they are capable of
meeting their rated capacity. These tests will demonstrate that the
batteries and the battery chargers are capable of performing their
mitigation function for analyzed accidents.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: William D. Reckley, Acting.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: November 16, 2001.
Description of amendment requests: The proposed amendments would
revise technical specification (TS) Table 3.3-4, ``Engineered Safety
Feature Actuation System Instrumentation Trip Setpoints.'' The proposed
changes are part of a planned design change to replace the existing 4kV
offsite power transformers, loss of voltage relays, and degraded
voltage relays with components of an improved design to increase the
reliability of offsite power for safety-related equipment.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
Probability of Occurrence of an Accident Previously Evaluated
The proposed changes to the degraded voltage and loss of voltage
setpoints and time delay affect when an emergency bus that is
experiencing low or degraded voltage will trip from offsite power
and shift to an emergency diesel generator. While the setpoints that
initiate this action will be modified, the function remains the
same. The setpoints have been analyzed to ensure spurious trips will
be avoided. The proposed changes will not significantly affect any
accident initiators or precursors. The format changes are intended
to improve readability, consistency with NUREG-1431, Revision 2, and
appearance. In addition, they do not alter any requirements. The
bases change provides explanatory information only. Thus, the
probability of occurrence of an accident previously evaluated is not
significantly increased.
Consequences of an Accident Previously Evaluated
The proposed changes to the degraded voltage and loss of voltage
setpoints and time delay affect when an emergency bus that is
experiencing low or degraded voltage will trip from offsite power
and shift to an emergency diesel generator. While the setpoints that
initiate this action will be modified, they are bounded by the
current safety analysis. The function of the plant equipment remains
the same. The proposed changes improve the reliability of safety-
related equipment to operate as designed. The format changes are
intended to improve readability, consistency with NUREG-1431,
Revision 2, and appearance. In addition, they do not alter any
requirements. The bases change provides explanatory information
only. Thus, the consequences of an accident previously analyzed are
not significantly increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes to the degraded voltage and loss of voltage
setpoints and time delay do not affect existing or introduce any new
accident precursors or modes of operation. The relays will continue
to detect undervoltage conditions and transfer safety loads to the
emergency diesel generators at a voltage level adequate to ensure
proper safety equipment performance and to prevent equipment damage.
The function of the relays remains the same. The format changes are
intended to improve readability, consistency with NUREG-1431,
Revision 2, and appearance. In addition, they do not alter any
requirements. The bases change provides explanatory information
only. Thus, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes will allow all safety-related loads to have
sufficient voltage to perform their intended safety function while
ensuring spurious trips are avoided. Thus, the results of the
accident analyses will not be affected as the input assumptions are
protected. The format changes are intended to improve readability,
consistency with NUREG-1431, Revision 2, and appearance. In
addition, they do not alter any requirements. The bases change
provides explanatory information only. Thus, the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: William D. Reckley, Acting Section Chief.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: August 2, 2001, as supplemented November
2, 2001.
Description of amendment request: The amendment would change the
Seabrook Station Technical Specification (TS) 6.15 to permit a one-time
exception to the 10-year frequency for the Integrated Leakage Rate Test
(ILRT). This exception would permit the existing ILRT frequency to be
extended from 10 years to 15 years from the last test completed on
October 30, 1992.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change to the Seabrook Station Technical
Specifications does not involve a significant increase in the
probability or consequences of an accident previously analyzed. The
proposed revision to TS 6.15 adds a one-time extension to the
current interval for the ILRT test. It is proposed that the current
test interval be extended from ten-years to fifteen-years from the
date of the last ILRT performed on October 30, 1992. The proposed
extension cannot increase the probability of an accident previously
evaluated since the test interval extension does not involve
modification of the plant, nor a operation of the plant that could
initiate an accident. The proposed extension of the ILRT does not
involve a significant increase in the consequences of an accident.
The increase in risk is very small because ILRTs identify only a few
potential leakage paths that cannot be identified by local leakage
rate [Type B and C]
testing, and the leaks that have been found by
ILRTs have been only marginally above existing requirements. An
analysis of the 144 ILRT results including 23 failures, found that
no ILRT failures were due to a containment liner breach. NUREG-1493
[``Performance-Based Containment Leak Test Program'']
concluded that
reducing the ILRT testing frequency to one per twenty years would
lead to an imperceptible increase in risk.
Therefore, it is concluded that the proposed change to TS 6.15
does not involve
[[Page 64299]]
a significant increase in the probability or consequence of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change to Technical Specification 6.15 does not
create the possibility of a new or different kind of accident from
any previously evaluated. The proposed change adds a one-time
extension to the current Integrated Leakage Rate Test frequency of
ten-years to fifteen-years from the date of the last test. The
proposed change cannot create the possibility of a new or different
type of accident since there are no physical changes being made to
the plant. Additionally, there are no changes to the operation of
the plant that could introduce a new failure mode creating an
accident.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed change does not involve a significant reduction in
the margin of safety. The proposed revision to TS 6.15 adds a one-
time extension to the current interval for the ILRT test. It is
proposed that the current test interval be extended from ten-years
to fifteen-years from the date of the last ILRT performed on October
30, 1992. A reduction in the ILRT frequency was found to lead to an
imperceptible decrease in the margin of safety. The estimated
increase in risk is very small because ILRTs identify only a few
potential leakage paths that cannot be identified by local leakage
rate [Type B and C]
testing, and the leaks that have been found by
ILRTs have been only marginally above existing requirements. A
Seabrook Station specific risk evaluation is consistent with the
generic conclusions identified in NUREG-1493.
Based on the above evaluation, North Atlantic concludes that the
proposed change to TS 6.15 does not constitute a significant hazard.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Section Chief: James W. Clifford.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: October 22, 2001.
Description of amendment request: The proposed amendment deletes
requirements from the technical specifications (TSs) (and, as
applicable, other elements of the licensing bases) to maintain a Post
Accident Sampling System (PASS). Licensees were generally required to
implement PASS upgrades as described in NUREG-0737, ``Clarification of
TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades was an outcome of the
lessons learned from the accident that occurred at TMI, Unit 2.
Requirements related to PASS were imposed by Order for many facilities
and were added to or included in the TSs for nuclear power reactors
currently licensed to operate. Lessons learned and improvements
implemented over the last 20 years have shown that the information
obtained from PASS can be readily obtained through other means or is of
little use in the assessment and mitigation of accident conditions.
The Nuclear Regulatory Commission (NRC) staff issued a notice of
opportunity for comment in the Federal Register on August 11, 2000 (65
FR 49271) on possible amendments to eliminate PASS, including a model
safety evaluation and model no significant hazards consideration (NSHC)
determination, using the consolidated line item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated October 22, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from
[[Page 64300]]
reactor accidents, results in a neutral impact to the margin of
safety. Methodologies that are not reliant on PASS are designed to
provide rapid assessment of current reactor core conditions and the
direction of degradation while effectively responding to the event
in order to mitigate the consequences of the accident. The use of a
PASS is redundant and does not provide quick recognition of core
events or rapid response to events in progress. The intent of the
requirements established as a result of the TMI-2 accident can be
adequately met without reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Section Chief: William D. Reckley, Acting.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: October 16, 2001.
Description of amendment request: The proposed amendment would
revise the Susquehanna Steam Electric Station (SSES), Units 1 and 2,
Technical Specifications (TSs). The licensee proposed to revise
selected sections of the administrative controls chapter of the TSs
consistent with Nuclear Regulatory Commission (NRC) approved Technical
Specification Task Force (TSTF) generic changes to NUREG-1433,
``Standard Technical Specifications for General Electric Plants (BWR/
4),'' Revision 1 (STS). The licensee also proposed editorial and
administrative changes to the affected sections.
The licensee categorized the proposed changes as either
``Administrative Changes'' or ``Less Restrictive Changes--Removed
Detail.'' The licensee categorized proposed changes consistent with the
approved versions of TSTF-273, TSTF-299, TSTF-308, TSTF-348, and TSTF-
364 as ``Administrative Changes.'' An administrative change involves
editorial restructuring of the current requirements, or modification of
wording that does not affect the technical content of the current TSs.
Administrative changes are not intended to add, delete, or relocate any
technical requirements of the current TSs. The licensee categorized
proposed changes consistent with the approved versions of TSTF-279 and
TSTF-363 as ``Less Restrictive Changes--Removed Detail.'' The proposed
changes involve moving details out of the TSs and into the TS Bases,
the updated Final Safety Analysis Report (UFSAR), the Technical
Requirements Manual (TRM), or other documents for which changes are
subject to regulatory control. The removal of this information is
considered to be less restrictive because it is no longer controlled by
the TS change process.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Administrative Changes
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
The proposed change involves reformatting, renumbering, and
rewording the existing [technical specification]
TS. The
reformatting, renumbering, and rewording process involves no
technical changes to the existing TS. As such, this change is
administrative in nature and does not affect the initiators of
analyzed events or assumed mitigation of accidents or transient
events. Therefore, this change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in methods governing normal plant operation. The proposed
change will not impose any new or eliminate any old requirements.
Thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed change will not reduce a margin of safety because
it has no effect on any safety analyses assumptions. Therefore, the
change does not involve a significant reduction in a margin of
safety.
Less Restrictive Changes--Removed Detail
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
The proposed change relocates certain details from the TS to
other documents under regulatory control. The TS Bases, [updated
final safety analysis report]
UFSAR, and [Technical Requirements
Manual]
TRM will be maintained in accordance with 10 CFR 50.59. In
addition to 10 CFR 50.59 provisions, the TS Bases are subject to the
change control provisions in the Administrative Controls Chapter of
the TS. The UFSAR is subject to the change control provisions of 10
CFR 50.71(e). Other documents are subject to controls imposed by TS
or regulations. Since any changes to these documents will be
evaluated, no significant increase in the probability or
consequences of an accident previously evaluated will be allowed.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in methods governing normal plant operation. The proposed
change will not impose any new or eliminate any old requirements,
and adequate control of the information will be maintained. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed change will not reduce a margin of safety because
it has no effect on any safety analyses assumptions. In addition,
the details to be moved from the TS to other documents are the same
as the existing TS. Since any future changes to these details will
be evaluated, no significant reduction in a margin of safety will be
allowed. A significant reduction in a margin of safety is not
associated with the elimination of the 10 CFR 50.92 requirement for
NRC review and approval of future changes to the relocated details.
The proposed change is consistent with NUREG 1433, issued by the NRC
staff, revising the TS to reflect the approved level of detail,
which indicates that there is no significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: L. Raghavan, Acting.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: October 25, 2001.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications (TSs) (and, as
applicable,
[[Page 64301]]
other elements of the licensing bases) to maintain a Post Accident
Sampling System (PASS). Licensees were generally required to implement
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three
Mile Island]
Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2. Requirements related to
PASS were imposed by an Order for many facilities and were added to or
included in the TS for nuclear power reactors currently licensed to
operate. Lessons learned and improvements implemented over the last 20
years have shown that the information obtained from PASS can be readily
obtained through other means or is of little use in the assessment and
mitigation of accident conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated October 25, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in [a]
Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in [a]
margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400
L Street, NW., Washington, DC 20005.
NRC Section Chief: L. Raghavan, Acting.
South Carolina Electric & Gas Company (SCE&G), South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station,
Unit No. 1, Fairfield County, South Carolina
Date of amendment request: August 20, 2001.
Description of amendment request: The licensee is proposing to
revise Virgil C. Summer Nuclear Station (VCSNS) Technical
Specifications (TS) to add a footnote to Table 3.3-3 regarding the
Steam Line Isolation and Engineered Safety Feature Actuation System
(ESFAS) functions. This revision will allow VCSNS to exclude ESFAS
steam line isolation instrumentation operability in Mode 3 when the
main steam isolation valves, along with associated bypass valves, are
closed and disabled, and ease the restriction of Specification 3.0.4
when performing reactor coolant system (RCS) resistance temperature
device (RTD) cross calibrations at temperatures below the ESFAS P-12
Interlock for Low-Low Tavg. This request is consistent in
part with the improved Standard Technical Specifications (ITS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
[[Page 64302]]
[The]
proposed changes involve upgrading the VCSNS TS to more
closely agree with ITS and does not result in any hardware changes.
The proposed change revises the applicability for the initiating
functions of the main steam isolation function such that when a main
steam line isolation valve is closed and the isolation function is
accomplished, the automatic initiation of this function is no longer
required to be operable. The ESFAS is not assumed to be an initiator
of any analyzed event. The role of the ESFAS is in mitigating and
thereby limiting the consequences of accidents. The proposed change
continues to adequately ensure the operability of the ESFAS main
steam line isolation function when the lines are unisolated and
thereby ensures the protection provided by the function remains
operable when required. The relaxation of the P-12 Function during
RCS RTD cross calibration allows all associated narrow range
temperature channels to remain in test, with test circuitry
installed, during the transition between Modes 4 and 3. Surveillance
performance is administratively controlled by plant procedures which
assure testing is conducted below the ESFAS P-12 interlock setpoint
of 552 deg.F and that TS limits for mode operability are not
exceeded. Therefore, the results of the analyses described in the
FSAR [Final Safety Analysis Report]
remain bounding. Additionally,
the proposed change does not impose any new safety analyses limits
or alter the plant's ability to detect or mitigate events.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes involve upgrading the ESFAS area of the
VCSNS TS to more closely agree with ITS and to support RCS RTD cross
calibration. The changes do not necessitate a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in parameters governing normal plant operation. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in margin of
safety?
The proposed change, which upgrades the ESFAS area of the VCSNS
TS to be more consistent with ITS and supports RCS RTD cross
calibration, does not have an adverse impact on any design basis
safety analysis. In combination with administrative controls,
required safety functions will continue to be accomplished in
accordance with safety analysis assumptions. As such, the results of
the analyses described in the FSAR remain bounding [, thus]
assuring
the proposed change does not involve a significant reduction in
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Section Chief: Richard Laufer, Acting.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: October 31, 2001.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications (TS) (and, as
applicable, other elements of the licensing bases) to maintain a Post
Accident Sampling System (PASS). Licensees were generally required to
implement PASS upgrades as described in NUREG-0737, ``Clarification of
TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades was an outcome of the
lessons learned from the accident that occurred at TMI, Unit 2.
Requirements related to PASS were imposed by Order for many facilities
and were added to or included in the TS for nuclear power reactors
currently licensed to operate. Lessons learned and improvements
implemented over the last 20 years have shown that the information
obtained from PASS can be readily obtained through other means or is of
little use in the assessment and mitigation of accident conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated October 31, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
[[Page 64303]]
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2 (SQN), Hamilton County, Tennessee
Date of amendment request: November 15, 2001 (TS-01-08).
Description of amendment request: The proposed amendment would
increase the full core thermal power rating by 1.3 percent from 3411
MWt to 3455 MWt, based on planned installation of the improved Caldon,
Incorporated (Caldon) Leading Edge Flow Meter, LEFMTM (LEFM)
feedwater flow measurement instrumentation. This change affects
Operating License Condition 2.C.(1) and Definition 1.26 for Rated
Thermal Power. In addition, changes are necessary to the reactor power
limits of Technical Specification (TS) Table 3.7.1 with an inoperable
main steam safety valve for both units and, for Unit 2 only, the
interval for which the pressure and temperature curves and the low
temperature over pressure protection curves (TS Figures 3.4-2, 3.4-3,
and 3.4-4) are valid. A change to the Bases for TS Section 3/4.7.1.1 is
also included to address the changes in main steam safety valve
capabilities.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The comprehensive analytical efforts performed to support the
proposed change included a review of the nuclear steam supply
systems (NSSSs) and components that could be affected by this
change. All systems and components will function as designed and the
applicable performance requirements have been evaluated and found to
be acceptable.
The primary loop components (reactor vessel, reactor internals,
control rod drive mechanism, loop piping and supports, reactor
coolant pump, steam generator and pressurizer) continue to comply
with their applicable structural limits and will continue to perform
their intended design functions. Thus, there is no increase in the
probability of a structural failure of these components. The rod
control cluster assembly (RCCA) drop time remains within the current
limits assumed in the accident analyses. Thus, there is no increase
in the consequences of the accidents which credit RCCA drop. Several
steam generator tubes may need to be plugged to preclude the
potential for U-bend fatigue if the plant operates below certain
steam pressure values. As long as these provisions are maintained,
there is no increase in the probability of an steam generator tube
rupture event. The leak before break analysis conclusions remain
valid and thus the limiting break sizes determined in this analysis
remain bounding.
All of the NSSS systems will continue to perform their intended
design functions during normal and accident conditions. The
pressurizer spray flow remains above its design value. Thus, the
control system design analyses that credit the spray flow do not
need to be modified for changes in this flow. The auxiliary systems
and components continue to comply with applicable structural limits
and will continue to perform their intended design functions. Thus,
there is no increase in the probability of a structural failure of
these components. All of the NSSS and/or balance of plant (BOP)
interface systems will continue to perform their intended design
functions. The steam generator safety valves will provide adequate
relief capacity to maintain the steam generators within design
limits. The steam dump system will still relieve 40 percent of the
maximum full load steam flow. The current loss-of-coolant accident
(LOCA) hydraulic forcing functions are still bounding. Thus, there
is no significant increase in the probability of an accident
previously evaluated.
The fuel has been completely analyzed to determine the effect of
the 1.3 percent power uprate. The fuel assembly and fuel rod
integrity have been evaluated. The change results in negligible
changes to the hydraulic lift forces and the existing holddown
margins remain acceptable. The increase in corrosion of the fuel
assembly structural Zircaloy-4 components due to a slight increase
in temperature is small, thus acceptable structural margin for
normal operating, faulted, and handling conditions exist. The fuel
assembly and fuel rod flow-induced vibration (FIV) performance
remains acceptable. The existing fuel assembly faulted condition
loading and analysis remain applicable and acceptable. The fuel rod
strain, creep collapse, and corrosion performance were evaluated at
the higher power level with acceptable results.
The fuel cycle design was evaluated and there was no significant
effect caused by the 1.3 percent power uprate. The operational
analysis of the core was evaluated for the change and found to
remain applicable with acceptable results.
The thermal-hydraulic analysis was evaluated and found to remain
applicable. The safety analysis addressed all Condition II, III, and
IV events with the conclusion that current analyses remain
applicable or bounding. The radiological consequences were evaluated
and determined to be bounded by current analyses.
Additionally, the current licensing basis steamline break and
LOCA mass and energy releases that are used to determine the peak
containment pressure and temperature limits continue to remain
bounding with the increase in power. Thus, there is no significant
increase in the consequences of an accident previously evaluated.
The heatup and cooldown curves for Unit 2 are now applicable for
14.5 EFPY [effective full-power year]
instead of 16 EFPY. The heatup
and cooldown curves define limits that still ensure the prevention
of nonductile failure for the SQN Units 1 and 2 reactor coolant
system (RCS). The design-basis events that were protected have not
changed. This modification does not alter any assumptions previously
made in the radiological consequence evaluations nor affect
mitigation of the radiological consequences of an accident described
in the Updated Final Safety Analysis Report. Therefore, the proposed
changes will not significantly increase the probability or
consequences of an accident previously evaluated.
The revised requirements for inoperable MSSVs [main steam safety
valves]
provide limits for the power range high flux trip
[[Page 64304]]
setpoint that ensure adequate relief capability for postulated
accidents. This change does not alter any plant systems, components,
or operating methods. Since the plant will continue to operate in
the same manner with the same protective features, this change will
not increase the possibility of an accident. The revised setpoint is
a conservative change that provides additional margin considering
the effect of the proposed power uprate. Since the revised setpoint
continues to provide an equivalent level of safety function, this
change will not significantly increase the consequences of an
accident and the offsite dose impact will not be significantly
increased.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
No new accident scenarios, failure mechanisms or single failures
are introduced as a result of the proposed changes. All systems,
structures, and components previously required for the mitigation of
an event remain capable of fulfilling their intended design
function. The proposed changes have no adverse effects on any
safety-related system or component and do not challenge the
performance or integrity of any safety-related system. Therefore, it
is concluded that the proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
Operation at the 3455 MWt core power does not involve a
significant reduction in a margin of safety. Extensive analyses of
the primary fission product barriers have concluded that all
relevant design criteria remain satisfied, both from the standpoint
of the integrity of the primary fission product barrier and from the
standpoint of compliance with the regulatory acceptance criteria.
The reduction in the EFPY for the Unit 2 heatup and cooldown curves
does not reduce the margin of safety since the curves define the
limits for ensuring the prevention of nonductile failure for the RCS
and these curves remain unchanged.
The pressure and temperature safety limits will be the same as
those for the current operating cycle, thus ensuring that the fuel
will be maintained within the same range of safety parameters that
form the basis for the Final Safety Analysis Report (FSAR) accident
evaluations.
The power uprate represents a small increase in the energy
production for the fuel cycle and is well within typical variations
that occur as a result of increases in cycle length and capacity
factor. The burnup of the fuel will increase proportionally with the
increase in power, but will not challenge the current licensed
burnup limit for Mark-BW fuel.
The slight increase in core average linear heat rate will result
in a slight loss of operating margin, but will not affect safety
margins. The centerline fuel melt and transient cladding strain
limits will not be affected by the power level uprate, but the
margin to these limits will decrease slightly. The LOCA FQ [power
peaking factor]
limits will not be altered since the increase in
core power is absorbed by reducing the power uncertainty used in
determination of the limits.
The power peaking limits that provide DNB [departure from
nucleate boiling]
protection are slightly lower resulting in a
proportional loss in DNB margins. The mechanical evaluation of the
fuel demonstrates that the power level uprate can be successfully
accomplished in compliance with all design criteria.
All FSAR Chapter 15 events have been evaluated and found to
remain applicable for the power uprate. The radiological
consequences analyses include an initial power assumption of 105
percent of 3411 MWt and remain bounding for the 1.3 percent power
uprate.
The more restrictive limits for the power range high flux trip
setpoint is based on calculations that ensure sufficient relief
capacity to meet accident mitigation requirements. This change will
appropriately limit reactor power levels, with inoperable MSSVs,
such that the margin of safety is maintained at an equivalent level
considering the proposed power uprate.
As appropriate, all evaluations have been performed using
methods that have either been reviewed and approved by the NRC or
that are in compliance with all applicable regulatory review
guidance and standards. All of the fuel and safety evaluations for
the 1.3 percent power uprate were performed with the Framatome-ANP
approved methodology listed in TS Section 6.9.1.14 of the SQN TSs.
Therefore, it is concluded that the proposed changes do not involve
a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request October 31, 2001.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications (TS) (and, as
applicable, other elements of the licensing bases) to maintain a Post
Accident Sampling System (PASS). Licensees were generally required to
implement PASS upgrades as described in NUREG-0737, ``Clarification of
TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades was an outcome of the
lessons learned from the accident that occurred at TMI, Unit 2.
Requirements related to PASS were imposed by Order for many facilities
and were added to or included in the TS for nuclear power reactors
currently licensed to operate. Lessons learned and improvements
implemented over the last 20 years have shown that the information
obtained from PASS can be readily obtained through other means or is of
little use in the assessment and mitigation of accident conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated October 31, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident
[[Page 64305]]
mitigation. Past experience has indicated that there exists in-plant
instrumentation and methodologies available in lieu of a PASS for
collecting and assimilating information needed to assess core damage
following an accident. Furthermore, the implementation of Severe
Accident Management Guidance (SAMG) emphasizes accident management
strategies based on in-plant instruments. These strategies provide
guidance to the plant staff for mitigation and recovery from a
severe accident. Based on current severe accident management
strategies and guidelines, it is determined that the PASS provides
little benefit to the plant staff in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: November 13, 2001.
Description of amendment request: The proposed amendment would
revise the Watts Bar Nuclear Unit 1 (WBN) Technical Requirements Manual
to add two new sections, TR 3.7.6, ``Shutdown Board Room (SDBR) Air
Conditioning System (ACS),'' and TR 3.7.7, ``Elevation 772.0 480 Volt
Board Room Air Conditioning (AC) Systems.'' Each section provides
specific actions and associated completion times for various out-of-
service conditions associated with the safety-related air conditioning
systems.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed revision to the WBN Technical Requirements Manual
(TRM) will provide formalized operational guidance for coping with
partial or complete unavailability of shut down board room (SDBR)
and 480V board room air conditioning (AC) equipment for limited
periods of time. The change does not impact the frequency of an
accident because failure of either the SDBR or the 480V board room
AC systems is not an initiator of any accident scenario. The change
does not modify any plant hardware including the air conditioning
systems, and none of their automatic control features or redundant
systems currently credited in failure analyses are being deleted,
modified, or otherwise replaced by operator actions as a result of
the proposed change.
The proposed TRM revision changes current plant operating
practice and WBN Final Safety Analysis Assumptions (FSAR)
assumptions by allowing continued power operation with both trains
of SDBR air conditioning concurrently inoperable and two 480V board
room AC systems of the same unit to be concurrently inoperable for a
limited duration, up to 12 hours. This condition is acceptable based
on the low probability of the occurrence of postulated accidents
resulting in core damage concurrent with multiple inoperable systems
or trains of cooling equipment during this timeframe, and based on
analyses which demonstrate that peak temperatures in each room
served by these systems remain below mild environment temperature
limits during this time period. Consequently, there is no
significant adverse impact on the ability of required safety-related
electrical equipment to continue to operate and perform their
required functions, during both normal operation and during design
basis events. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not modify any plant hardware including
the subject air conditioning systems. The change provides specific
operational guidance for coping with partial or complete
unavailability of shut down board room and 480V board room air
conditioning equipment. No new accident or event initiators are
created by allowing multiple air conditioning systems to be
unavailable for the limited time period of 12 hours. The supported
electrical equipment remains capable of performing its intended
function both during normal operations and post accident. Therefore,
the proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed TRM revision changes current FSAR assumptions by
allowing continued power operation with both trains of SDBR air
conditioning concurrently inoperable and allowing two 480V board
room air conditioning systems of the same unit to be inoperable for
a limited duration, up to 12 hours. This condition does not
significantly reduce the margin of safety due to the low probability
of the occurrence of a postulated accident resulting in core damage
concurrent with multiple inoperable systems or trains of cooling
equipment during the limited time period. In addition, transient
temperature analyses demonstrate that peak temperatures in each room
served by these systems remain below mild environment temperature
limits for a period of 24 hours assuming a complete loss of air
conditioning to all rooms served by the SDBR and 480V board room AC
systems concurrently. The analysis is bounding for normal
operational
[[Page 64306]]
conditions. Consequently, there is no significant adverse impact on
the ability of required safety-related electrical equipment to
continue to operate and perform their required functions during both
normal operation and during design basis events. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: October 2, 2001.
Brief description of amendments: The proposed amendment deletes
requirements from the Technical Specifications (TSs) (and, as
applicable, other elements of the licensing bases) to maintain a Post
Accident Sampling System (PASS). Licensees were generally required to
implement PASS upgrades as described in NUREG-0737, ``Clarification of
TMI [Three Mile Island Nuclear Station]
Action Plan Requirements,'' and
Regulatory Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear
Power Plants to Assess Plant and Environs Conditions During and
Following an Accident.'' Implementation of these upgrades was an
outcome of the lessons learned from the accident that occurred at TMI,
Unit 2 (TMI-2). Requirements related to PASS were imposed by Order for
many facilities and were added to or included in the TSs for nuclear
power reactors currently licensed to operate. Lessons learned and
improvements implemented over the last 20 years have shown that the
information obtained from PASS can be readily obtained through other
means or is of little use in the assessment and mitigation of accident
conditions. The Nuclear Regulatory Commission (NRC) staff issued a
notice of opportunity for comment in the Federal Register on August 11,
2000 (65 FR 49271), on possible amendments to eliminate PASS, including
a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on October 31, 2000 (65 FR 65018).
The licensee affirmed the applicability of the following NSHC
determination in its application dated October 2, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents, and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan, the
emergency operating procedures, and site survey monitoring that
support modification of emergency plan protective action
recommendations.
Therefore, the elimination of PASS requirements from the TSs
(and other elements of the licensing bases) does not involve a
significant increase in the consequences of any accident previously
evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: October 25, 2001.
[[Page 64307]]
Brief description of amendments: The proposed amendment would
revise Technical Specification (TS) 4.2.1, ``Fuel Assemblies,'' for
Comanche Peak Steam Electric Station (CPSES) Units 1 and 2 to allow the
use of ZIRLOTM test assemblies and to further allow, `` * *
* A limited number of lead test assemblies * * * be placed in non-
limiting core regions.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Changing the technical specifications within limits of the
bounding accident analyses cannot change the probability of an
accident previously evaluated, nor will it increase radiological
consequences predicted by the analyses of record. Controlling the
use of lead test assemblies according to limitations approved by the
NRC [Nuclear Regulatory Commission]
constrains fuel performance
within limits bounded by existing design basis accident and
transient analyses.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Inclusion in the reactor core of lead test assemblies according
to limitations set by the NRC for lead test assemblies and of a
design approved by the NRC ensures that their effect on core
performance remains within existing design limits. Use of fuel
assemblies whose design has been previously approved by the NRC as
lead test assemblies is consistent with current plant design bases,
does not adversely affect any fission product barrier, and does not
alter the safety function of safety significant systems, structures
and components or their roles in accident prevention or mitigation.
Currently licensed design basis accident and transient analyses of
record remain valid.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not alter the manner in which Safety
Limits, Limiting Safety System Setpoints, or Limiting Conditions for
Operation are determined. This proposed clarification of TS 4.2.1 is
bounded by existing limits on reactor operation. It leaves current
limitations for use of lead test assemblies in place, conforms to
plant design bases, is consistent with current safety analyses, and
limits actual plant operation within analyzed and licensed
boundaries.
Therefore, the proposed change does not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: November 8, 2001.
Brief description of amendments: The amendments would add the
following to the Technical Specifications (TSs) for Comanche Peak Steam
Electric Station (CPSES): (1) the phrase, ``* * * or if open, capable
of being closed * * *'' to the TS Limiting Condition for Operation
3.9.4 for the equipment hatch, during core alterations or movement of
irradiated fuel assemblies inside containment; and (2) the requirement
to verify the capability to install the equipment hatch in a new
Surveillance Requirement (SR) 3.9.4.2. Nothing is proposed to be
deleted from the TSs. Existing SR 3.9.4.2 would be renumbered SR
3.9.4.3, but would not otherwise be changed. Item (1) will allow the
equipment hatch to be open during the conditions stated above.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes will allow the equipment hatch to be open
during CORE ALTERATIONS and movement of irradiated fuel assemblies
inside containment. The status of the equipment hatch during
refueling operations has no affect on the probability of the
occurrence of any accident previously evaluated. The proposed
revision does not alter any plant equipment or operating practices
in such a manner that the probability of an accident is increased.
Since the consequences of a fuel handling accident inside
containment with an open equipment hatch are bounded by the current
analysis described in the FSAR [Final Safety Analysis Report]
and
the probability of an accident is not affected by the status of the
equipment hatch, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not create any new failure modes for any
system or component, nor do they adversely affect plant operation.
No new equipment will be added and no new limiting single failures
will be created. The plant will continue to be operated within the
envelope of the existing safety analysis.
Therefore, the proposed changes do not create a new or different
kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The previously determined radiological dose consequences for a
fuel handling accident inside containment with the personnel air
lock doors open remain bounding for the proposed changes. These
previously determined dose consequences were determined to be well
within the limits of 10 CFR [Part]
100 and they meet the acceptance
criteria of SRP [Standard Review Plan]
section 15.7.4 and GDC
[General Design Criterion]
19.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: November 7, 2001.
Description of amendment request: A change is proposed to Technical
specification 3.0.3 to allow a longer period of time to perform a
missed surveillance. The time is extended from the current limit of ``
* * * up to 24 hours or up to the limit of the specified Frequency,
whichever is less'' to `` * * * up to 24 hours or up to the limit of
the specified Frequency, whichever is greater.'' In addition, the
following requirement would be added to the specification: ``A risk
evaluation shall be performed for any Surveillance
[[Page 64308]]
delayed greater than 24 hours and the risk impact shall be managed.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments
concerning missed surveillances, including a model safety evaluation
and model no significant hazards consideration (NSHC) determination,
using the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on September 28, 2001 (66 FR 49714). The licensee affirmed the
applicability of the following NSHC determination in its application
dated November 7, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated The proposed change relaxes the time allowed to
perform a missed surveillance. The time between surveillances is not
an initiator of any accident previously evaluated. Consequently, the
probability of an accident previously evaluated is not significantly
increased. The equipment being tested is still required to be
operable and capable of performing the accident mitigation functions
assumed in the accident analysis. As a result, the consequences of
any accident previously evaluated are not significantly affected.
Any reduction in confidence that a standby system might fail to
perform its safety function due to a missed surveillance is small
and would not, in the absence of other unrelated failures, lead to
an increase in consequences beyond those estimated by existing
analyses. The addition of a requirement to assess and manage the
risk introduced by the missed surveillance will further minimize
possible concerns. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO [Limiting Condition for
Operation]
is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on the margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards considerations.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: November 7, 2001 (ULNRC-04557).
Description of amendment request: The proposed amendment would
revise Surveillance Requirements (SRs) 3.3.1.2 and 3.3.1.3 in the
Technical Specifications (TSs) on reactor trip system (RTS)
instrumentation. The proposed change to SR 3.3.1.2 would replace the
reference to the nuclear instrumentation system (NIS) channel output by
a reference to the power range channel output, and delete Note 1 to the
SR. The change to SR 3.3.1.3 is editorial in nature.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since there are
no hardware changes. The RTS instrumentation will be unaffected.
Protection systems will continue to function in a manner consistent
with the plant design basis. All design, material, and construction
standards that were applicable prior to the request are maintained.
The probability and consequences of accidents previously
evaluated in the FSAR [Final Safety Analysis Report]
are not
adversely affected because the change to the NIS power range channel
daily surveillance assures the conservative response of the channel
even at part-power levels.
The proposed changes modify the NIS power range channel daily
surveillance requirement to assure the NIS power range functions are
tested in a manner consistent with the safety analysis and licensing
basis.
The proposed changes will not affect the probability of any
event initiators. There will be no degradation in the performance
of, or an increase in the number of challenges imposed on, safety-
related equipment assumed to function during an accident situation.
There will be no change to normal plant operating parameters or
accident mitigation performance.
The proposed changes will not alter any assumptions or change
any mitigation actions in the [accident]
radiological consequence
evaluations in the FSAR.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. This amendment will not affect the normal method of plant
operation or change any operating parameters. No performance
requirements or response time limits will be affected; however, the
proposed TS Bases changes impose explicit NIS power range high trip
setpoint adjustment requirements prior to adjusting indicated power
in a decreasing power direction. These requirements are consistent
with assumptions made in the safety analysis and licensing basis.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this amendment. There will be no adverse effect or challenges
imposed on any safety-related system as a result of this amendment.
[[Page 64309]]
This amendment does not alter the design or performance of the
7300 Process Protection System, Nuclear Instrumentation System, or
Solid State Protection System used in the plant protection systems.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes require a revision to the criteria for
implementation of NIS power range channel adjustments based on
secondary power calorimetric calculations; however, the changes do
not eliminate any RTS surveillances or alter the frequency of
surveillances required by the Technical Specifications. The revision
to the criteria for implementation of the daily surveillance will
have a conservative effect on the performance of the NIS power range
channels, particularly at part-power conditions. The nominal trip
setpoints specified in the Technical Specification Bases and the
safety analysis limits assumed in the transient and accident
analyses are unchanged. None of the acceptance criteria for any
accident analysis is changed.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on the overpower
limit, departure from nucleate boiling ratio (DNBR) limits, heat
flux hot channel factor (FQ), nuclear enthalpy rise hot
channel factor (FH), loss of coolant accident peak cladding
temperature (LOCA PCT), peak local power density, or any other
margin of safety. The radiological dose consequence acceptance
criteria listed in the Standard Review Plan will continue to be met.
The imposition of appropriate surveillance testing requirements
will not reduce any margin of safety since the changes will assure
that safety analysis assumptions on equipment operability are
verified on a periodic frequency.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Virginia Electric and Power Company, Docket No. 50-280, Surry Power
Station, Unit No. 1, Surry County, Virginia
Date of amendment request: October 15, 2001, as supplemented
November 8, 2001.
Description of amendment request: The proposed amendment would
revise Technical Specifications Section 4.4. The proposed changes would
permit a one-time 5-year extension of the 10-year performance-based
Type A test interval established in NEI 94-01, ``Nuclear Energy
Institute Industry Guideline for Implementing Performance-Based Option
of 10 CFR Part 50, Appendix J,'' Revision 0, July 26, 1995.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed extension to Type A testing cannot increase the
probability of an accident previously evaluated since extension of
the containment Type A testing is not a physical plant modification
that could alter the probability of accident occurrence nor, is an
activity or modification by itself that could lead to equipment
failure or accident initiation.
The proposed extension to Type A testing does not result in a
significant increase in the consequences of an accident as documented
in NUREG-1493. The NUREG notes that very few potential containment
leakage paths are not identified by Type B and C tests. It concludes
that reducing the Type A (ILRT) testing frequency to once per twenty
years leads to an imperceptible increase in risk.
Surry provides a high degree of assurance through indirect testing
and inspection that the containment will not degrade in a manner
detectable only by Type A testing. The last two Type A tests identified
containment leakage within acceptance criteria, indicating a very leak-
tight containment. Inspections required by the ASME Code are also
performed in order to identify indications of containment degradation
that could affect leak-tightness. Also, maintaining the containment
subatmospheric during operations provides constant monitoring of the
leaktightness of the containment structure. Separately, Type B and C
testing, required by Technical Specifications, identifies any
containment opening from design penetrations, such as valves, that
would otherwise be detected by a Type A test. These factors establish
that an extension to the Surry Type A test interval will not represent
a significant increase in the consequences of an accident.
2. Does the proposed license amendment create the possibility of a
new or different kind of accident from any accident previously
evaluated?
The proposed revision to Technical Specifications adds a one-time
extension to the current interval for Type A testing for Surry Unit 1.
The current test interval of ten years, based on past performance,
would be extended on a one-time basis to fifteen years from the last
Type A test. The proposed extension to Type A testing does not create
the possibility of a new or different type of accident since there are
no physical changes being made to the plant and there are no changes to
the operation of the plant that could introduce a new failure.
3. Does the proposed amendment involve a significant reduction in a
margin of safety?
The proposed revision to Surry Technical Specifications adds a one-
time extension to the current interval for Type A testing. The current
test interval of ten years, based on past performance, would be
extended on a one-time basis to fifteen years from the last Type A test
for Surry Unit 1. The proposed extension to Type A testing will not
significantly reduce the margin of safety. The NUREG-1493 generic study
of the effects of extending containment leakage testing found that a
20-year interval in Type A leakage testing resulted in an imperceptible
increase in risk to the public. NUREG-1493 found that, generically, the
design containment leakage rate contributes about 0.1 percent of the
overall risk and that decreasing the Type A testing frequency would
have a minimal [effect]
on this risk since 95% of the Type A detectable
leakage paths would already be detected by Type B and C testing. In
addition, the risk impact on the total integrated (fifteen year total)
Surry Unit 1 plant risk above baseline, for those accident sequences
influenced by Type A testing, is only 0.004%. Furthermore, for Surry,
maintaining the containment subatmospheric during plant operations
further reduces the risk of any containment leakage path going
undetected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams,
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia
23219.
NRC Section Chief: Richard J. Laufer, Acting.
[[Page 64310]]
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: May 31, 2001 as supplemented October 17,
2001.
Description of amendment request: The proposed changes would revise
the Technical Specifications and associated Bases to provide a separate
allowed outage time for the backup air supply for the pressurizer
power-operated relief valves (PORVs).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Dominion has reviewed the requirements of 10 CFR 50.92 as they
relate to the proposed change for Surry Units 1 and 2 and determined
that a significant hazards consideration is not involved. The
following is provided to support this conclusion.
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change does not introduce any new mechanisms for
the initiation of transients or accidents or for the failure of
equipment relied upon in the accident analyses to mitigate the
consequences of accidents. The impact of the proposed change on the
availability and reliability of the pressurizer PORVs is negligible.
Therefore the accident analysis results and conclusions remain
bounding.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated.
There are no modifications to the plant as a result of the
changes. No new accident or event initiators are created by changing
the required actions for various conditions of PORV inoperability.
The proposed change will not introduce any new equipment failure
modes that could initiate accidents or change the analysis results
presented in the UFSAR [Updated Final Safety Analysis Report].
3. Does the change involve a significant reduction in a margin
of safety.
The proposed change will not alter the limiting results of the
safety analyses presented in Chapter 14 of the UFSAR. Provision of
an allowed outage time for the pressurizer PORV backup air system
and of more condition specific and appropriate actions for various
types of PORV inoperability has an insignificant impact on the
availability and reliability of the PORVs for performing their
safety related functions.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams,
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia
23219.
NRC Section Chief: Richard J. Laufer, Acting.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management Systems (ADAMS) Public Electronic
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/
NRC/ADAMS/index.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by e-mail to pdr@nrc.gov.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: July 26, 2001.
Brief description of amendments: The amendments modify Technical
Specifications 5.5.14.b and 5.5.14.b.2, Technical Specification Bases
Control Program, such that they are consistent with Title 10 of the
Code of Federal Regulations (10 CFR 50.59).
Date of issuance: November 21, 2001.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 247 and 222.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 5, 2001 (66
FR 46475) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated November 21, 2001.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: April 19, 2001.
Brief description of amendment: The amendment changes the River
Bend Station Technical Specifications (TSs) to allow an increase in the
number of spent fuel assemblies (SFAs) to be stored in the spent fuel
pool from the current TS limit of 2680 SFAs to 3104 SFAs.
Date of issuance: November 19, 2001.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 123.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 18, 2001 (66 FR
52948) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 19, 2001.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 23, 2001, as supplemented by letter
dated October 25, 2001.
Brief description of amendment: The change deletes Technical
Specification
[[Page 64311]]
(TS) 3.9.12, ``Fuel Handling Building Ventilation System,'' and TS
3.3.3.1 Surveillance Requirements for the Fuel Storage Pool area
radiation monitors.
Date of issuance: November 21, 2001.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 176.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 22, 2001 (66 FR
44169). The October 25, 2001, supplement contained clarifying
information that did not change the scope of the July 23, 2001,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 21, 2001.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: August 22, 2001.
Brief description of amendments: The amendments revise the
Technical Specifications for St. Lucie Units 1 and 2 to allow small,
controlled, safe insertions of positive reactivity while in shutdown
modes.
Date of Issuance: November 19, 2001.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 179 and 122.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 19, 2001 (66
FR 48287).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 19, 2001.
No significant hazards consideration comments received: No.
GPU Nuclear Inc., Docket No. 50-320, Three Mile Island Nuclear Station,
Unit 2, Dauphin County, Pennsylvania
Date of amendment request: June 21, 2001.
Brief description of amendment request: The amendment revises Three
Mile Island Nuclear Station, Unit 2 Technical Specifications
Administrative Controls section to provide consistency with the changes
to the revised subsection 50.59 of Title 10 of the Code of Federal
Regulations, as published in the Federal Register on October 4, 1999
(64 FR 53582).
Date of issuance: November 28, 2001
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 57.
Facility Operating License No. DPR-73: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 31, 2001 (66 FR
55020).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 28, 2001.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, BerrienCounty, Michigan
Date of application for amendments: May 15, 2001.
Brief description of amendments: The amendments change TS 3/
4.8.2.2, ``A. C. Distribution Shutdown,'' TS 3/4.8.2.4 ``D. C.
Distribution--Shutdown,'' and TS 3/4.9.4, ``Containment Building
Penetrations.'' The proposed amendments replaces the current required
actions in TSs 3/4.8.2.2. and 3/4.8.2.4, to establish containment
integrity within 8 hours if less than the specified minimum complement
of A.C. or D.C. busses and equipment is operable in Modes 5 and 6 with
new actions which require to immediately suspend operations involving
core alterations, positive reactivity changes, and movement of
irradiated fuel assemblies, to immediately initiate actions to restore
the required busses and return equipment to operable status, and to
immediately declare the associated required residual heat removal
loop(s) inoperable. The proposed new actions are consistent with
NUREG--1431, ``Standard Technical Specifications, Westinghouse
Plants,'' Revision 1.
In addition, the proposed amendments will change TS 3/4.9.4 to add
options to use containment penetration closure methods that are
equivalent to those that are currently required by the TSs during core
alterations or movement of irradiated fuel in containment, and to allow
unisolation of some penetrations under administrative control.
Date of issuance: November 21, 2001.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 259 and 242.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 12, 2001 (66 FR
31709).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 21, 2001.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: October 8, 2001.
Brief description of amendments: The amendments revised the
Technical Specifications to allow the main control room boundary to be
opened intermittently under administrative controls and to allow 24
hours to restore the main control room boundary to Operable status
before requiring the plant to perform an orderly shutdown.
Date of issuance: November 26, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 225 and 168.
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 26, 2001 (66 FR
54301).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 26, 2001.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: September 7, 2001 (TS 01-09).
Brief description of amendment: The amendment revised Technical
Specifications (TS) Section 3.6.11, ``Ice Bed,'' Surveillance
Requirement (SR) 3.6.11.2, SR 3.6.11.3, and the associated Bases, to
lower the minimum average ice basket weight from 1236 pounds to 1110
pounds.
Date of issuance: November 29, 2001.
Effective date: As of the date of its issuance and shall be
implemented no later than Mode 4 during startup from Cycle 4 refueling
outage.
Amendment No.: 33.
[[Page 64312]]
Facility Operating License No. NPF-90: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 17, 2001 (66 FR
52804).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 29, 2001.
No significant hazards consideration comments received: No.
Note: The publication date for this notice will change from
every other Wednesday to every other Tuesday, effective January 8,
2002. The notice will contain the same information and will continue
to be published biweekly.
Dated at Rockville, Maryland, this 3rd of December, 2001.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 01-30455 Filed 12-11-01; 8:45 am]
BILLING CODE 7590-01-P