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Pennsylvania Power Company, Ohio Edison Company, FirstEnergy Nuclear Operating Company, Beaver Valley Power Station, Unit Nos. 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: March 15, 2001 (Volume 66, Number 51)]
[Notices]
[Page 15147-15149]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr15mr01-105]

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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-334 and 50-412]


Pennsylvania Power Company, Ohio Edison Company, FirstEnergy
Nuclear Operating Company, Beaver Valley Power Station, Unit Nos. 1 and
2; Environmental Assessment and Finding of No Significant Impact

    The U.S. Nuclear Regulatory Commission (NRC) is considering
issuance of an amendment to Facility Operating License Nos. DPR-66 and
NPF-73, issued to FirstEnergy Nuclear Operating Company, et al. (FENOC,
the licensee), for operation of the Beaver Valley Power Station (BVPS),
Unit Nos. 1 and 2, located in Shippingport, Pennsylvania.

Environmental Assessment

Identification of the Proposed Action

    The proposed action would authorize revisions to the BVPS Updated
Final Safety Analysis Reports (UFSARs) involving calculated doses and
associated descriptions/information for selected Design Basis Accidents
(DBAs). The following DBAs were revised as documented in the licensee's
submittals for the BVPS, Unit 1 UFSAR (Exclusion Area Boundary (EAB)
doses are calculated over the first 2 hours following the accident and
all other doses are calculated over the duration of the accident).

Loss of Offsite AC Power

    Changes include revisions to Table 14.1-3 to reflect corrected or
conservative analysis input parameter values or input assumptions based
on plant design and operation. The analysis methodology remained the
same as had been previously reviewed and approved by the NRC for BVPS,
Unit 1, and the revised analysis resulted in no increase in calculated
doses.

Fuel-Handling Accident (FHA)

    Changes include revisions to Section 14.2.1 and Tables 14.2-6 and
14.2-6a to reflect corrected or conservative analysis input parameter
values or input assumptions based on plant design and operation. The
analysis methodology remained the same as had been previously reviewed
and approved by the NRC for BVPS, Unit 1. Because the FHA dose analysis
takes credit for removal of organic iodine by the supplemental leak
collection and release system (SLCRS), the licensee added a safety
factor of  2 in accordance with guidance given in Generic
Letter (GL) 99-02, ``Laboratory Testing of Nuclear-Grade Activated
Charcoal.'' GL 99-02 guidance included testing nuclear-activated
charcoal filters to a more stringent requirement (supported by the
safety factor) than that assumed in the safety analysis to
conservatively account for potential degradation to nuclear-grade
charcoal filters over the surveillance interval. As a consequence of
this safety factor, the calculated doses increased. The calculated
thyroid dose at the EAB increased from 14.6 rem to 24.6 rem. The
calculated control room operator thyroid dose increased from 3.2 rem to
6.26 rem. These doses are well within the applicable DBA dose
guidelines set forth in Title 10 of the Code of Federal Regulations (10
CFR) Section 100.11 (EAB thyroid dose of 300 rem from iodine exposure)
and 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 19
(control room operator whole body dose of 5 rem or its equivalent to
any organ).

Accidental Release of Waste Gas

    Changes include revisions to Section 14.2.3 and Table 14.2-8 to
reflect corrected or conservative analysis input parameter values or
input assumptions based on plant design and operation. Some changes to
the analysis methodology were made. As a result of the revisions to the
analysis, the calculated control room whole body dose increased from
less than .01 rem to .0295 rem.

Steam Generator Tube Rupture (SGTR)

    Changes include revisions to Section 14.2.4 and Table 14.2-9 to
reflect corrected or conservative analysis input parameter values or
input assumptions based on plant design and operation. The methodology
for the offsite dose analysis was changed to that of the current SGTR
analysis of record for the control room operator dose. As a result, the
calculated thyroid dose at the EAB for the coincident iodine spike
increased from .9 rem to 1.37 rem.

Rod Cluster Control Assembly Ejection

    Changes include revisions to Table 14.2.12 to reflect corrected or
conservative analysis input parameter values or input assumptions based
on plant design and operation. The analysis methodology remained the
same as had been previously approved by the NRC for BVPS, Unit 1. The
revised analysis

[[Page 15148]]

showed no increase in any calculated doses.

Single Reactor Coolant Pump Locked Rotor

    Changes include revisions to Section 14.2.7 and Table 14.2-4b to
reflect corrected or conservative analysis input parameter values or
input assumptions based on plant design and operation. In addition, the
coincident iodine spike, previously assumed to occur, is removed from
the analysis, based on the assumption of 18-percent failed fuel. In its
previous analysis of record, the licensee assumed both the coincident
iodine spike and 18-percent failed fuel. SRP 15.3.3 guidance encourages
the use of either of the assumptions but not both. The 18-percent
failed fuel assumption is more conservative than the iodine spike
occurrence assumption because the calculated dose consequences
resulting from assuming 18-percent failed fuel are more severe than the
calculated dose consequences resulting from the iodine spike
occurrence. The revised analysis showed no increase in any calculated
doses.

Loss of Reactor Coolant from Small Ruptured Pipes/Loss-of-Coolant
Accidents (LOCA)

    Changes include revisions to Section 14.3.5 and Tables 14.3-10,
14.3-13, and 14.3-14a to reflect corrected or conservative analysis
input parameter values or input assumptions based on plant design and
operation. In addition, some analysis methodology was revised. Shine
from the area beneath the control room that is not within the control
room ventilation envelope was added as an additional contributor to the
control room dose. Also, because the LOCA dose analysis takes credit
for removal of organic iodine by the SLCRS, the licensee added a safety
factor of  2 in accordance with the guidance given in GL 99-
02. As a result of the changes to the LOCA dose analysis, the
calculated control room whole body dose increased from .17 rem to .71
rem.
    The following DBAs were revised as documented in the licensee's
submittals for the BVPS, Unit 2 UFSAR.

Steam System Piping Failures (Main Steam Line Break Accident)

    Changes include revisions to Section 15.1.5 and Table 15.1-3 to
reflect corrected or conservative analysis input parameter values or
input assumptions based on plant design and operation. The analysis
methodology remained the same as had been previously reviewed and
approved by the NRC for BVPS, Unit 2. The revised analysis showed no
increase in any calculated doses.

Loss of AC Power

    Changes include revisions to Section 15.2.6 and Table 15.2-2 to
reflect corrected or conservative analysis input parameter values or
input assumptions based on plant design and operation. The analysis
methodology remained the same as had been previously reviewed and
approved by the NRC for BVPS, Unit 2. The revised analysis showed no
increase in any calculated doses.

Reactor Coolant Pump Shaft Seizure

    Changes include revisions to Section 15.3.3 and Table 15.3-3 to
reflect corrected or conservative analysis input parameter values or
input assumptions based on plant design and operation. Unlike the
previous analysis of record, isolation of the control room was not
assumed to occur for the revised analysis. The control room isolation
function remains operationally unchanged. It is conservatively not
credited in the analysis. As a result, the calculated control room
operator thyroid dose increased from 1.7 rem to 7.46 rem. This is well
within the 10 CFR Part 50, Appendix A, GDC 19 DBA dose guidelines for
control room operators.

Rod Cluster Control Assembly Ejection

    Changes include revisions to Section 15.4.8 and Table 15.4-3 to
reflect corrected or conservative analysis input parameter values or
input assumptions based on plant design and operation. The analysis
methodology remained the same as had been previously reviewed and
approved by the NRC for BVPS, Unit 2. The revised analysis showed no
increase in any calculated doses.

Failure of Small Lines Carrying Primary Coolant Outside Containment

    Changes include revisions to Section 15.6.2 and Table 15.6-2 to
reflect corrected or conservative analysis input parameter values or
input assumptions based on plant design and operation. The analysis
methodology remained the same as had been previously reviewed and
approved by the NRC for BVPS, Unit 2. The revised analysis showed no
increase in any calculated doses.

Steam Generator Tube Rupture

    Changes include revisions to Section 15.6.3 and Table 15.6-5b to
reflect corrected or conservative analysis input parameter values or
input assumptions based on plant design and operation. The analysis
methodology remained the same as had been previously reviewed and
approved by the NRC for BVPS, Unit 2. The revised analysis showed no
increase in any calculated doses.

Loss-of-Coolant Accidents

    Changes include revisions to Section 15.6.5 and Tables 15.6-11 and
15.6-12 to reflect corrected or conservative analysis input parameter
values or input assumptions based on plant design and operation. The
analysis methodology remained the same as had been previously reviewed
and approved by the NRC for BVPS, Unit 2. As a result of the revisions,
the calculated control room operator whole body dose increased from .32
rem to .33 rem and the calculated control room operator thyroid dose
increased from 1.3 rem to 2 rem.

Waste Gas System Failures

    Changes include revisions to Section 15.7.1 and Tables 15.7-1 and
15.7-2 to reflect corrected or conservative analysis input parameter
values or input assumptions based on plant design and operation. The
analysis methodology remained the same as had been previously reviewed
and approved by the NRC for BVPS, Unit 2. The revised analysis showed
no increase in any calculated doses.
    The proposed action is in accordance with the licensee's
application for amendment dated May 12, 2000, as supplemented on June
19, November 2, and December 1, 2000 and January 29, 2001.

The Need for the Proposed Action

    The proposed revisions are a result of an extensive review by the
licensee to assess the dose calculations' input parameter values, input
assumptions, design basis consistency, calculation methodologies, and
conservatism.
    The change is not the result of hardware changes to the plant or a
change in operating practices. The proposed changes reflect corrected
or conservative analysis input parameters, assumptions, and new
analysis methodologies. In addition, some changes were made in response
to GL 99-02.

Environmental Impacts of the Proposed Action

    The NRC has completed its evaluation of the proposed action and
concludes that the assumptions and methodologies used by the licensee
in the analyses are acceptable and that there is reasonable assurance,
in the event of a postulated DBA, that the calculated offsite doses
would continue to be well within the 10 CFR part 100 guidelines, and
the calculated control room operator doses would continue to be less
than the 10

[[Page 15149]]

CFR part 50, Appendix A, GDC 19 guidelines.
    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 are no significant nonradiological environmental
impacts associated with the proposed action.
    Accordingly, the NRC concludes that there are no significant
environmental impacts associated with the proposed action.

Alternatives to the Proposed Action

    As an alternative to the proposed action, the staff considered
denial of the proposed action (i.e., the ``no-action'' alternative).
Denial of the application would result in no change in current
environmental impacts. The environmental impacts of the proposed action
and the alternative action are similar.

Alternative Use of Resources

    This action does not involve the use of any resources not
previously considered in the Final Environmental Statement for the
Beaver Valley Power Station, Unit Nos. 1 and 2.

Agencies and Persons Consulted

    In accordance with its stated policy, on February 1, 2000, the
staff consulted with the Pennsylvania State official, Mr. L. Ryan, of
the Pennsylvania Department of Environmental Protection Bureau,
Division of Nuclear Safety, 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 concludes
that the proposed action will not have a significant effect on the
quality of the human environment. Accordingly, the NRC 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 May 12, 2000, as supplemented on June 19,
November 2, and December 1, 2000, and January 29, 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 on
the NRC Web site, http:\\www.nrc.gov (the Electronic Reading Room).

    Dated at Rockville, Maryland, this 9th day of March 2001.

    For the Nuclear Regulatory Commission.
Lawrence J. Burkhart,
Project Manager, Section 1, Project Directorate I, Division of
Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 01-6405 Filed 3-14-01; 8:45 am]
BILLING CODE 7590-01-P

 
 

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