[Federal Register Volume 66, Number 51 (Thursday, March 15, 2001)]
[Notices]
[Pages 15147-15149]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-6405]


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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