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STP Nuclear Operating Company; South Texas Project Electric Generating Station, Units 1 and 2; Environmental Assessment and Finding of No Significant Impact
Note: EPA no longer updates this information, but it may be useful as a reference or resource.
[Federal Register: June 14, 2001 (Volume 66, Number 115)]
[Notices]
[Page 32397-32399]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr14jn01-137]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-498 and 50-499]
STP Nuclear Operating Company; South Texas Project Electric
Generating Station, Units 1 and 2; Environmental Assessment and Finding
of No Significant Impact
The U.S. Nuclear Regulatory Commission (NRC) is considering
issuance of exemptions from certain regulations found in 10 CFR parts
21, 50, and 100 for Facility Operating License Nos. NPF-76 and NPF-80,
issued to STP Nuclear Operating Company (STPNOC or the licensee) for
operation of the South Texas Project Generating Station, Units 1 and 2,
(STP) located in Matagorda County, Texas.
Environmental Assessment
Identification of the Proposed Action
The proposed action would grant the licensee relief from certain
special treatment requirements found in Title 10 of the Code of Federal
Regulations, Parts 21, 50, and 100 (10 CFR parts 21, 50 and 100) for
certain structures, systems, and components (SSCs). The licensee has
used a risk-informed process to categorize SSCs as low safety
significant (LSS) or non-risk significant (NRS); and other SSCs as
medium safety significant (MSS) or high safety significant (HSS). The
purpose of this categorization process is to identify those SSCs for
which the special treatment requirements may be relaxed. Currently, LSS
and NRS SSCs, which are not as risk significant as MSS and HSS SSCs,
are treated with the same level of protection. The licensee is seeking
limited exemptions from the following regulations for just those SSCs
that have been categorized as LSS or NRS:
1. Requirements for quality assurance (QA) found in:
a. 10 CFR part 50, Appendix B, ``Quality Assurance Criteria for
Nuclear Power Plants and Fuel Reprocessing Plants,'' for QA
requirements on SSCs that are safety-related (with the exception of the
Criterion III, ``Design Control,'' Criterion XV, ``Nonconforming
Materials, Parts, or Components,'' and Criterion XVI, ``Corrective
Action'');
b. 10 CFR part 50, Appendix A, General Design Criteria (GDC) 1,
``Quality Standards and Records,'' for SSCs important to safety that
contains quality assurance program and record keeping requirements;
c. 10 CFR 50.34(b)(6)(ii) that requires the licensee to describe in
the Final Safety Analysis Report how 10 CFR part 50, Appendix B,
requirements are being satisfied;
d. 10 CFR 50.54(a)(3) regarding NRC review and approval of changes
to the QA program that result in a reduction in commitments in the
program description as accepted by the NRC for LSS and NRS SSC program
descriptions; and,
e. 10 CFR 21.3 defining the term ``basic component'' that includes
safety-related LSS and NRS SSCs and impose 10 CFR part 21 requirements
for procurement, dedication, and reporting.
2. Requirements for environmental qualification (EQ) found in:
a. 10 CFR 50.49(b) that defines the scope of electric components
important to safety subject to the EQ program requirements of 10 CFR
50.49;
b. 10 CFR part 50, Appendix A, GDC 2, ``Design Bases for Protection
Against Natural Phenomena,'' for tests and inspections to demonstrate
that SSCs important to safety are designed to withstand the effects of
natural phenomena without loss of capability to perform their safety
functions;
c. 10 CFR part 50, Appendix A, GDC 4, ``Environmental and Dynamic
Effects Design Bases,'' for tests and inspections to demonstrate that
SSCs important to safety are able to withstand environmental conditions
of normal operation, maintenance, testing, and postulated accidents;
and,
d. 10 CFR part 100, Appendix A, Sections VI(a)(1) and (a)(2) for
testing and inspection to demonstrate that SSCs within the scope of
these regulations\1\ are designed to remain functional during a safe-
shutdown earthquake and operating-basis earthquake, respectively, and
10 CFR 50.34(b)(10) and 10 CFR 50.34(b)(11) to the extent that they
reference the 10 CFR part 100, Appendix A, criteria, discussed above.
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\1\ The scope of (a)(1) are those SSCs necessary to assure (i)
the integrity of the reactor coolant pressure boundary, (ii) the
capability to shut down the reactor and maintain it in a safe
condition, or continued (iii) the capability to prevent or mitigate
the consequences of accidents which could result in potential
offsite exposures comparable to the guideline exposures of part 100.
The scope of (a)(2) are those SSCs necessary for continued operation
without undue risk to the health and safety of the public.
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3. Requirements for testing and inspection found in:
a. 10 CFR part 50, Appendix A, GDC 18, ``Inspection and Testing of
Electric Power Systems,'' that requires SSCs important to safety be
designed to permit inspection and testing; and
b. 10 CFR part 50, Appendix J, Option B, section III.B, ``Type B
and C Tests,'' that requires Type C containment
[[Page 32398]]
isolation valve leak rate tests for safety-related SSCs.
4. Requirements for monitoring the effectiveness of maintenance
under 10 CFR 50.65 for safety-related SSCs and nonsafety-related SSCs
that are relied upon to mitigate accidents or transients or are used in
plant emergency operating procedures, or whose failure could prevent
safety-related SSCs from fulfilling their safety-related function, or
whose failure could cause a reactor scram or actuation of a safety-
related system. The licensee is requesting an exemption to exclude the
LSS and NRS SSCs from the scope of the maintenance rule but would still
conduct monitoring at the plant, system, or train level. Failure of an
LSS or NRS SSC would not count as a Maintenance Rule Functional Failure
unless the failure caused a failure of a high or medium safety
significant function.
5. Industry code standards found in:
a. 10 CFR 50.55a(f) and (g) that require repair and replacement,
inservice testing (IST), and inservice inspection (ISI), under Section
XI of the American Society of Mechanical Engineers Boiler and Pressure
Vessel Code; and,
b. 10 CFR 50.55a(h) that imposes the quality and qualification
requirements of Sections 4.3 and 4.4 of Institute of Electrical and
Electronics Engineers (IEEE) 279, ``Criteria for Protection Systems for
Nuclear Power Plant Generating Stations,'' for electric SSCs important
to safety.
6. 10 CFR 50.59 to the extent that this regulation requires a
written evaluation and prior NRC review and approval of changes in
special treatment requirements for LSS and NRS SSCs.
The proposed action is in accordance with the licensee's
application for exemption dated July 13, 1999, as supplemented on
October 14 and 22, 1999; January 26 and August 31, 2000; and January
15, 18, and 23, March 19, and May 8 and 21, 2001 (hereinafter, the
submittal).
The Need for the Proposed Action
The exemptions are necessary to provide the licensee relief from
regulatory requirements found in 10 CFR parts 21, 50, and 100 for LSS
and NRS components currently within the scope of these regulations. In
accordance with 10 CFR 21.7 and 10 CFR 50.12, the Commission may grant
exemptions from the requirements of 10 CFR parts 21 and 50,
respectively, under certain circumstances. Further, the NRC staff has
determined that the requested exemptions from 10 CFR part 100, Appendix
A, sections VI(a)(1) and (a)(2) may be granted in accordance with the
requirements of 10 CFR 50.12. The NRC staff has approved a Graded
Quality Assurance Program for STP. Exemptions from certain special
treatment requirements are necessary to realize the full benefit of the
Graded Quality Assurance Program.
The exemption is also necessary to reduce occupational radiation
exposures and costs that would be expended in providing qualifications,
quality assurance controls, maintenance, monitoring requirements,
testing, and inspections for the LSS and NRS components that may not be
necessary to maintain safety.
Environmental Impacts of the Proposed Action
The NRC staff has completed its evaluation of the proposed action
and concludes that many of the exemption requests are not necessary
for, or are not consistent with the objective of, maintaining design
and functionality of an SSC and will not be granted. The NRC staff has
determined that some of the exemption requests that would remove LSS
and NRS SSCs from the scope of the regulations, if granted, would not
present an undue risk to the public health and safety. The regulations
for which exemptions are to be granted are listed below and are
referred to as the proposed action in the following sections.
a. 10 CFR 21.3--definition of basic component;
b. 10 CFR 50.34(b)(10) and 10 CFR 50.34(b)(11), impose the
requirements of 10 CFR part 100, Appendix A, section VI(a)(1) and (2);
c. 10 CFR 50.49(b), scope of electrical equipment subject to
environmental qualification requirements [design aspects of 10 CFR
50.49(e)(1) through (7) continue to apply];
d. 10 CFR 50.55a(f)--IST requirements;
e. 10 CFR 50.55a(g)--repair and replacement, and ISI requirements;
f. 10 CFR 50.55a(h), quality and qualification requirements of
sections 4.3 and 4.4 of IEEE 279;
g. 10 CFR 50.59--written evaluations and prior NRC review and
approval for changes to special treatment requirements;
h. 10 CFR 50.65(b)--scope of maintenance rule [the requirements of
10 CFR 50.65(a)(4) continue to apply];
i. 10 CFR part 50, App. B--quality assurance requirements (the
requirements of Criteria III, ``Design Control,'' XV, ``Nonconforming
Materials, Parts, or Components,'' and XVI, ``Corrective Action,''
continue to apply);
j. 10 CFR part 50, Appendix J, Option B, section III.B, Type C
containment isolation valve leak rate tests only;
k. 10 CFR part 100, Appendix A, sections VI(a)(1) and (2), seismic
requirements for safe shutdown and operating basis earthquakes.
The regulations, listed above, apply to SSCs that are located
entirely within the restricted area and, if the exemptions are granted,
would not result in off-site impacts due to normal operation. The NRC
staff evaluated the licensee's probabilistic risk analysis (PRA)
sensitivity study that addressed the overall impact of reduced
treatment for LSS and NRS SSCs on plant risk for those LSS SSCs that
are modeled in the STP PRA.\2\ Since the impact on failure rates for
these SSCs resulting from a reduction in special treatment requirements
is not known, a factor of 10 increase in the failure rates of all LSS
SSCs modeled in the STP PRA was used. The results of the sensitivity
analysis showed that the overall plant risk for a core damage event
increased by 2.7 percent. The large early release frequency increased
by about 1.2 percent. The NRC staff finds the sensitivity study to be
an acceptable method of ensuring that the cumulative risk is only
slightly impacted when predicting significant changes in the SSC
failure rates, which may not occur. Therefore, the NRC staff finds that
the postulated change in failure rates of the LSS SSCs that are modeled
in the PRA would be expected to have a low overall impact on plant
risk.
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\2\ There are no NRS SSCs and limited LSS SSCs modeled in the
plant's PRA due to a negligible impact on risk or due to implicit
modeling.
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On the other hand, the proposed exemptions may have a beneficial
impact on occupational exposure, since the additional requirements for
QA, EQ, monitoring, testing, and inspection for certain LSS and NRS
components would not be necessary. The magnitude of this benefit has
not been quantified.
The proposed action will not significantly increase the probability
or consequences of accidents, no changes are being made in the types of
any effluents that may be released off site, and there is no
significant increase in occupational or public radiation exposure.
Therefore, there are no significant radiological environmental impacts
associated with the proposed action.
With regard to potential nonradiological impacts, the proposed
action does not involve any historic sites. It does not affect
nonradiological plant effluents and has no other environmental impact.
Therefore, there
[[Page 32399]]
are no significant nonradiological environmental impacts associated
with the proposed action.
Accordingly, the NRC staff concludes that there are no significant
environmental impacts associated with the proposed action.
Alternatives to the Proposed Action
As an alternative to the exemption requests listed above, the NRC
staff considered denial of the proposed action (i.e., the ``no-action''
alternative). Denial of the application would result in no significant
change in current environmental impacts.
Another alternative is to await applicable regulations that are the
result of a future rulemaking under Option 2 of the Commission's
alternatives to risk inform 10 CFR part 50 of the NRC's regulations
discussed in SECY-98-300, ``Options for Risk-Informed Revisions to 10
CFR part 50, ``Domestic Licensing of Production and Utilization
Facilities'.'' The exemptions requested by the licensee are a proof-of-
concept for this broader rulemaking effort. The Commission plans to use
the STPNOC exemption request and other industry pilot programs to
assist with the development of the revised risk-informed 10 CFR part
50. The only adverse environmental impact associated with this proposed
action would be a slight increase in the risk of an accident, but this
impact would not be significantly changed with the alternative of
awaiting a rulemaking. Therefore, any relief granted under a subset of
a larger set of risk-informed regulations under Option 2 in lieu of the
exemption requests would not provide a significant benefit to public
health or safety, or the environment. The environmental impacts
associated with granting the exemptions found to be acceptable by the
NRC staff and the alternatives listed above are similar.
Alternative Use of Resources
This action does not involve the use of any resources not
previously considered in the Final Environmental Statement (NUREG-1171)
for the South Texas Project, Units 1 and 2, dated August 1986.
Agencies and Persons Consulted
In accordance with its stated policy, on June 1, 2001, the NRC
staff consulted with the Texas State official, Arthur C. Tate, of the
Division of Compliance and Inspection, Bureau of Radiation Control,
Texas Department of Health, regarding the environmental impact of the
proposed action. The State official had no comments.
Finding of No Significant Impact
On the basis of the environmental assessment, the NRC staff
concludes that the proposed action will not have a significant effect
on the quality of the human environment. Accordingly, the NRC staff has
determined not to prepare an environmental impact statement for the
proposed action.
For further details with respect to the proposed action, see the
licensee's letter dated July 13, 1999, as supplemented on October 14
and 22, 1999; January 26 and August 31, 2000; and January 15, 18, and
23, March 19, and May 8 and 21, 2001. Documents may be examined and/or
copied for a fee, at the NRC's Public Document Room, located at One
White Flint North, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible electronically
from the ADAMS Public Library component of the NRC web site http://
www.nrc.gov (Electronic Reading Room).
Dated at Rockville, Maryland, this 8th day of June, 2001.
For the Nuclear Regulatory Commission.
Cynthia A. Carpenter,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 01-14976 Filed 6-13-01; 8:45 am]
BILLING CODE 7590-01-P