ML061370696

Vermont Yankee - Attachment 5 to the NRC Staff'S Answer to New England Coalition'S Request for Leave to File New Contentions

ML061370696
Person / Time
Site:	Vermont Yankee
Issue date:	05/01/2006
From:	NRC/OGC
To:	Atomic Safety and Licensing Board Panel
Byrdsong A T
References
50-271-OLA, ASLBP 04-832-02-OLA, RAS 11605
Download: ML061370696 (26)

Text

L May 1,2006 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of 1 1

ENTERGY NUCLEAR VERMONT YANKEE, ) Docket No. 50-271-0LA LLC and ENTERGY NUCLEAR 1 OPERATIOIVS, INC. ) ASLBP NO.04-832-02-OLA

)

(Vermont Yankee Nuclear Power Station) )

NRC STAFF'S ANSWER TO NEW ENGLAND COAI-ITION'S REQUEST FOR LEAVE TO FILE NEW CONTEN-I-IONS

OFFICE O F NUCLEAR REACTOR REGULATION REVIEW STANDARD FOR EXTENDED POWER UPRATES APPROVED BY: /RA/

L. Marsh, Director Division of Licensing Project Management Office of Nuclear Reactor Regulation CONTACT: Mohammed A. Shuaibi, NRR (301) 415-2859 mas4@nrc.gov RS-001, Revision 0 DECEMBER 2003

RS-001, REVISION 0 REVIEW STANDARD FOR EXTENDED POWER UPRATES TABLE OF CONTENTS PURPOSE BACKGROUND GUIDANCE SECTION 1 PROCEDURAL GUIDANCE 1.1 - Processing Extended Power Uprate Applications Figure 1.I-1 EPU Process Flow Chart SECTION 2 -TECHNICAL REVIEW GUIDANCE 2.1 - Reviewing Extended Power Uprate Applications Matrix 1 - Materials and Chemical Engineering Matrix 2 - Mechanical and Civil Engineering Matrix 3 - Electrical Engineering Matrix 4 - lnstrumentation and Controls Matrix 5 - Plant Systems Matrix 6 - Containment Review Considerations Matrix 7 - Habitability, Filtration, and Ventilation Matrix 8 - Reactor Systems Matrix 9 - Source Terms and Radiological Consequences Analyses Matrix 10 - Health Physics Matrix 11 - Human Performance Matrix 12 - Power Ascension and Testing Plan Matrix 13 - Risk Evaluation SECTION 3 DOCUMENTATION OF REVIEW 3.1 - Documenting Reviews of Extended Power Uprate Applications 3.2 - Boiling-Water Reactor Template Safety Evaluation lnsert 1 - Materials and Chemical Engineering lnsert 2 Mechanical and Civil Engineering lnsert 3 - Electrical Engineering lnsert 4 - lnstrumentation and Controls lnsert 5 - Plant Systems lnsert 6 - Containment Review Considerations lnsert 7 - Habitability, Filtration, and Ventilation lnsert 8 - Reactor Systems lnsert 9 - Source Terms and Radiological Consequences Analyses lnsert 10 - Health Physics lnsert 11 - Human Performance lnsert 12 - Power Ascension and Testing Plan lnsert 13 - Risk Evaluation 3.3 - Pressurized-Water Reactor Template Safety Evaluation lnsert 1 - Materials and Chemical Engineering lnsert 2 - Mechanical and Civil Engineering lnsert 3 - Electrical Engineering lnsert 4 - lnstrumentation and Controls lnsert 5 - Plant Systems lnsert 6 - Containment Review Considerations lnsert 7 - Habitability, Filtration, and Ventilation lnsert 8 - Reactor Systems lnsert 9 - Source Terms and Radiological Consequences Analyses lnsert 10 - Health Physics lnsert 11 - Human Performance lnsert 12 - Power Ascension and Testing Plan lnsert 13 - Risk Evaluation SECTION 4 INSPECTION GUIDANCE 4.1 - Inspection Requirements DECEMBER 2003

RS-001, REVISION 0 REVIEW STANDARD FOR EXTENDED POWER UPRATES PURPOSE The purpose of this review standard is to provide guidance for the Nuclear Regulatory Commission (NRC) staff's review of extended power uprate (EPU) applications to enhance consistency, quality, and completeness of reviews.

This review standard also informs licensees of the guidance documents 'the staff uses when reviewing EPU applications. These documents provide acceptance criteria for the areas of review. This should allow licensees to prepare EPU applications that are complete with respect to the areas that are within the staff's scope of review. To further improve the efficiency of the staff's review of EPU applications, licensees are encouraged to provide, with their EPU applications, markups of the matrices in Section 2.1 and template safety evaluation inserts in Section 3 of this review standard to identify any differences between the information in the review standard and the design bases of their plants.

Use of this review standard should not undermine the MRC1slongstanding topical report review and approval process. If a licensee references an NRC-approved topical report for an area covered by this review standard, the staff will review the application only to ensure that the licensee is applying the topical report under conditions for which the topical report was approved, using appropriate plant-specific inputs.

The staff will review plants against their design bases. Licensees are encouraged to provide, with their EPU applications, markups of the matrices in Section 2.1 and template safety evaluation inserts in Section 3 of this review standard to identify any differences between the information in the review standard and the design bases of their plants. This should help the staff identify areas where the criteria and/or guidance in the review standard does not apply to the plant under review. The staff does not intend to impose the criteria and/or guidance in this review standard on plants whose design bases do not include these criteria and/or guidance.

No backfitting is intended or approved in connection with the issuance of this review standard.

In addition to this review standard, the NRC maintains a Web site on power uprates at http://www.nrc.~ov/reactors/o~eratin~/licensin/~ower-u~rates.html. Some of the material on this Web site includes:

the status of completed, ongoing, and expected power uprate reviews general guidance related to power uprates references to publicly available correspondence related to reviews of recently completed power uprates (including licensees' responses to NRC staff requests for additional information, as well as NRC staff safety evaluations)

DECEMBER 2003

RS-001, REVISION 0 REVIEW STANDARD FOR EXTENDED POWER UPRATES BACKGROUND Facility operating licenses and technical specifica.tions specify the maximum power level at which commercial nuclear power plants may be operated. NRC approval is required for any changes to facility operating licenses or technical specifications. The process for making changes to facility operating licenses and technical specifications is governed by Title 10 of the Code of Federal Regulations, Part 50.

The process of increasing the licensed power level at a corr~mercialnuclear power plant is called a "power uprate." Power uprates are categorized based on the magnitude of the power increase and the methods used to achieve the increase. Measurement uncertainty recapture power uprates result in power level increases that are less than 2 percent and are achieved by implementing enhanced techniques for calculating reactor power. Stretch power uprates typically result in power level increases that are up to 7 percent and do not generally involve major plant modifications. EPUs result in power level increases that are greater than stretch power uprates and usually require significant modifications to major plant equipment. The NRC has approved EPUs for increases as high as 20 percent. This review standard is applicable to EPUs.

This review standard establishes standardized review guidance and acceptance criteria for the staff's reviews of EPU applications to enhance the consistency, quality, and completeness of reviews. It serves as a tool for the staff's use when processing EPU applications in that it provides detailed references to various NRC documents containing information related to the specific areas of review.

This review standard also informs licensees of the guidance documents the staff will use when reviewing EPU applications. This will help licensees prepare EPU applications that address those topics necessary for a complete application. By addressing the areas in the review standard, a licensee could prepare and submit a more complete application and thus minimize the staff's need for requests for additional information (RAls). This would improve the efficiency of the staff's reviews.

The development of this review standard included an evaluation of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (SRP), to determine the applicability and adequacy of the various SRP sections to the review of EPU applications and development/revision of guidance, as necessary. During this evaluation, the staff considered the versions of the SRP sections identified in the matrices in Section 2 of this review standard. To determine the need for guidance beyond that in the SRP, the staff reviewed: ( I ) safety evaluations for previously approved power uprates, (2) previously approved topical reports for EPUs, (3) various reports related to lessons learned from the Maine Yankee experience (e.g., Report of the Maine Yankee Lessons Learned Task Group, dated December 1996), and (4) generic communications. The staff also considered feedback from internal and external stakeholders. In addition, the staff reviewed RAls issued for recent EPU applications to ensure that the review standard adequately addresses areas where repeat RAls have been issued.

The staff reviewed NRC procedural gtiidance d~curnenisto identify those applicable to processing EPU applications. The review of these documents also included consideration of the recommendations in various reports related to the Maine Yankee experience and the feedback received from internal and external stakeholders.

Figure 1 provides a graphical representation of the development of the review standard.

DECEMBER 2003

RS-001, REVISION 0 REVIEW STANDARD FOR EXTENDED POWER UPRATES GUIDANCE This review standard provides guidance for processing EPU applications (Section 1) performing technical reviews (Section 2) preparing safety evaluations to document the reviews (Section 3)

This review standard also includes a reference to the NRC's Inspection Manual, which provides guidance for conducting inspections related to the implementation of power uprates (Section 4).

DECEMBER 2003

for BOILING-WATER REACTOR XTNDD POWR UPRATE

RS-001,REVISION 0 REVIEW STANDARD FOR EXTENDED POWER UPRATES SECTION 3.2 BWR TEMPLATE SAFETY EVALUATION TABLE OF CONTENTS

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 1.1 Ap~lication . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2 Backqround . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3 Licensee'sAD~roach. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1.4 Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 1.5 Method of NRC Staff Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0 EVALUA-TION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1 Materials and Chemical Enqineerinq . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . &

2.2 Mechanical and Civil Enqineerinq . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 2.3 Electrical Enqineerinq . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 2.4 Instrumentation and Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5 Plant Svstems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 2.6 Containment Review Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . +

2.7 Habitabilitv, Filtration, and Ventilation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 2.8 Reactor Svstems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.9 Source Terms and Radioloqical Consequences Analvses . . . . . . . . . . . . . . . . . . +

2.1 0 Health Phvsics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 2.1 1 Human Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A 2.1 2 Power Ascension and Testinq Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . +

2.1 3 Risk Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 3.0 FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATION CHANGES . . A 4.0 REGULATORY COMMITMENTS ......................................... +

5.0 RECOMMENDED AREAS FOR INSPECTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -&

6.0 STATE CONSULTATION

7.0 ENVlROlUlVlElVTAL CONSIDERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -&

Attachment:

List of Acronyms

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO ANIEIVDIWEWT NO. TO FACILITY OPERATING LICENSE NO. TXXX-XXI

[NAME OF LICENSEE]

JNAME OF FACILITY]

DOCKET NO. 50-TXXXI 1.1 Application By application dated [ 1, as supplemented by letter[s] dated [ 1, the [Name of Licensee]

(the licensee) requested changes to the Facility Operating License and Technical Specifications (TSs) for the [Plant Name]. The supplemental letter[s] dated [ 1, provided additional clarifying information that did not expand the scope of the initial application and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on [date] (XX FR XXXX).

The proposed changes would increase the maximum steady-state reactor core power level from [current licensed power level] megawatts thermal (MWt) to [power level proposed b y the licensee] MWt, which is an increase of approximately [#I percent. The proposed increase in power level is considered an extended power uprate (EPU).

1.2 Backsround

[Plant Name] is a boiling-water reactor (BWR) plant of the BWR/[#] design with a Mark-[#]

containment. [Plant Name] has the following special featureslunique designs:

[Insert any special featureslunique designs]

The NRC originally licensed [Plant Name] on [date] for operation at [original licensed power level] MWt. [By Amendment No. [###I dated [ 1, the NRC granted a power uprate t o

[Plant Name] of [##I percent, allowing the plant to be operated at [current licensed power level] MWt.] Therefore, the proposed EPU would result in an increase of approximately

[##I percent over the original licensed power level [and [##I percent over the current licensed power level] for [Plant Name].]

SECTION 3.2 BWR TEMPLATE SAFETY EVALUATION DECEMBER ZOO3

1.3 Licensee's Approach The licensee's application for the proposed EPU follows the guidance in the Office of Nuclear Reactor Regulation's (NRR's) Review Standard (RS)-001, "Review Standard for Extended Power Uprates," to the extent that the review standard is consistent with the design basis of the plant. Where differences exist between the plant-specific design basis and RS-001, the licensee described the differences and provided evaluations consistent with the design basis of the plant. The licensee also used [Identify topical reports or other documents used by the licensee for guidance related to the scope of the proposed EPU; NRC staff approvals, ranges of applicability, any Iimitationslrestrictions associated with the documents; and consistency of the licensee's application with the ranges of applicability and limitationslrestrictions. The discussion in this section is to cover topical reports and other documents referenced for the overall power uprate process. It is not intended to cover topical reports and other documents for specific methods of analyses. Topical reports and other documents referenced for specific methods of analyses are t o be covered in the applicable technical evaluation section of this safety evaluation].

lnsert this sentence if the licensee is planning to implement the EPU in one stage.

['The licensee plans to implement the EPU in one step. The licensee plans t o make the modifications necessary t o implement the EPU during the refueling outage i n

[season year (e.g., fall 2003)l. Subsequently, the plant will be operated atI#[ MWt starting in Cycle [##].I lnsert this paragraph if the licensee is planning to implement the EPU in stages:

['The licensee plans to implement the EPU in [#] steps of [## and ##] percent. The licensee plans to make modifications necessary to implement the first step during the refueling outage in [season year (e.g., fall 2003)l. Subsequently, the plant will be operated at [##I MWt during Cycle [##I. 'The remainder of the modifications will be completed during the refueling outage in [season year (e.g., fall 2003)], with subsequent operation atI#[ MWt starting in Cycle # I[#

.]

1.4 Plant Modifications The licensee has determined that several plant modifications are necessary to implement the proposed EPU. The following is a list of these modifications and the licensee's proposed schedule for completing them.

[Provide a list of plant modifications.]

The NRC staff's evaluation of the licensee's proposed plant modifications is provided in Section 2.0 of this safety evaluation.

SECTION 3.2 BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003

1.5 Method of NRC Staff Review The NRC staff reviewed the licensee's application to ensure that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) activities proposed will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. The purpose of the NRC staff's review is to evaluate the licensee's assessment of the impact of the proposed EPU on design-basis analyses. The WRC staff evaluated the licensee's application and supplements. The NRC staff also evaluated [Include additional review items, as necessary (e.g., audits of certain information at the plant and vendor sites, and independent analyses), f o r areas where such analyses were deemed appropriate by the NRC staff].

In areas where the licensee and its contractors used NRC-approved or widely accepted methods in performing analyses related to the proposed EPU, the NRC staff reviewed relevant material to ensure that the licensee/contractor used the methods consistent with the limitations and restrictions placed on the methods. In addition, the NRC staff considered the affects of the changes in plant operating conditions on the use of these methods to ensure that the methods are appropriate for use at the proposed EPU conditions. Details of the NRC staff's review are provided in Section 2.0 of this safety evaluation.

Audits of analyses supporting the EPU were conducted in relation to the following topics:

[Provide a list of areas for which audits were performed.]

The results of the audits are discussed in section 2.0 of this safety evaluation.

Independent NRC staff calculations were performed in relation to the following topics:

[Provide a list of areas for which independent NRC staff calculations were performed.]

The results of the calculations are discussed in section 2.0 of this safety evaluation.

2.0 EVALUATION 2.1 Materials and Chemical Enaineerinq SEE INSERT IFOR SECTION 3.2 OF RS-001 2.2 Mechanical and Civil Enaineerinq SEE INSERT 2 FOR SECTION 3.2 OF RS-001 2.3 Electrical Ensineerinq SEE INSERT 3 FOR SECTION 3.2 OF RS-001 SECTION 3.2 BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003

2.4 Instrumentation and Controls SEE INSERT 4 FOR SECTION 3.2 OF RS-001 2.5 Plant Svstems SEE INSERT 5 FOR SECTION 3.2 OF RS-001 2.6 Containment Review Considerations SEE INSERT 6 FOR SECTION 3.2 OF RS-001 2.7 Habitabilitv, Filtration, and Ventilation SEE INSERT 7 FOR SECTION 3.2 OF RS-001 2.8 Reactor Svstems SEE INSERT 8 FOR SECTION 3.2 OF RS-001

/'

2.9 Source Terms and Radioloaical Conseauences Analvses SEE INSERT 9 FOR SECTION 3.2 OF RS-001

\

2.10 Health Phvsics SEE INSERT 10 FOR SECTION 3.2 OF RS-001 2.1 1 Human Performance SEE INSERT 11 FOR SECTION 3.2 OF RS-001 2.12 Power Ascension and Testins Plan SEE INSERT 12 FOR SEC'TION 3.2 OF RS-001 2.13 Risk Evaluation SEE INSERT 13 FOR SECTION 3.2 OF RS-001 3.0 FACILITY OPERATllUG LICENSE AND TECHNICAL SPECIFICATION CHANGES To achieve the EPU, the licensee proposed the following changes to the Facility Operating License and TSs for [Plant Name].

[Provide a list of license and TSs changes (including license conditions) and an NRC staff evaluation o f each.]

SECTION 3.2 - BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003

4.0 REGULATORY COMMITMENTS Insert the following sentence if the licensee has not made any regulatory commitments in support of the EPU.

The licensee has made no regulatory commitments i n its application for the EPU.

Insert the following if the licensee has made regulatory commitments in support of the EPU.

The licensee has made the following regulatory commitment(s):

[Provide a summary of each regulatory commitment made by the licensee.]

The NRC staff finds that reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the above regulatory commitment(s) are best provided by the licensee's administrative processes, including its commitment management program. The above regulatory commitments do not warrant the creation of regulatory requirements (items requiring prior NRC approval of subsequent changes).

5.0 RECOMMENDED AREAS FOR INSPECTION As described above, the NRC staff has conducted an extensive review of the licensee's plans and analyses related to the proposed EPU and concluded that they are acceptable. The NRC staff's review has identified the following areas for consideration by the NRC inspection staff during the licensee's implementation of the proposed EPU. These areas are recommended based on past experience with EPUs, the extent and unique nature of modifications necessary to implement the proposed EPU, and new conditions of operation necessary for the proposed EPU. They do not constitute inspection requirements, but are intended to give inspectors insight into important bases for approving the EPU.

[Provide list of recommended areas for inspection.]

6.0 STATE CONSULTATION

In accordance with the Commission's regulations, the [Name of State] State official was notified of the proposed issuance of the amendment. The State official had [no] comments.

[If comments were received, address them here.]

7.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, 51.33, and 51.35, a draft Environmental Assessment and finding of no significant impact was prepared and published in the Federal Register on

[Date] ( FR ). The draft E~ivironmentalAssessment provided a 30-day opportunity for public comment. If no comments were received, use the following sentence: [No comments were received on the draft Environmental Assessment.] If comments were received, use the following sentence: [The NRC staff received comments which were addressed in the final environmental assessment.] The final Environmental Assessment was published in the Federal Register on [Date] ( FR ). Accordingly, based upon the environmental SECTION 3.2 BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003

assessment, the Commission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.

8.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

9.0 REFERENCES

1. RS-001, Revision 0, "Review Standard for Extended Power Uprates," December 2003.

2. [Insert additional references as necessary]

Attachment:

List of Acronyms Principal Contributors:

Date:

SECTION 3.2 BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003

INSERT 8 POR SCT10N 3.2 BWR TMPlAT S A P N VALUATION

2.9 Source Terms and Radioloqical Conseauences Analvses 2.9.1 Source Terms for Radwaste Svstems Analvses Requlatow Evaluation The NRC staff reviewed the radioactive source term associated with EPUs to ensure the adequacy of the sources of radioactivity used by the licensee as input to calculations to verify that the radioactive waste management systems have adequate capacity for the treatment of radioactive liquid and gaseous wastes. The NRC staff's review included the parameters used to determine (1) the concentration of each radionuclide in the reactor coolant, (2) the fraction of fission product activity released to the reactor coolant, (3) concentrations of all radionuclides other than fission products in the reactor coolant, (4) leakage rates and associated fluid activity of all potentially radioactive water and steam systems, and (5) potential sources of radioactive materials in effluents that are not considered in the plant's [Updated Safety Analysis Report or Updated Final Safety Analysis Report] related to liquid waste management systems and gaseous waste management systems. The NRC's acceptance criteria for source terms are based on (1) 10 CFR Part 20, insofar as it establishes requirements for radioactivity in liquid and gaseous effluents released to unrestricted areas; (2) 10 CFR Part 50, Appendix I, insofar as it establishes numerical guides for design objectives and limiting conditions for operation to meet the "as low as is reasonably achievable" criterion; and (3) GDC-60, insofar as it requires that the plant design include means to control the release of radioactive effluents. Specific review criteria are contained in SRP Section 11.I.

Technical Evaluation

[Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]

Conclusion The NRC staff has reviewed the radioactive source term associated with the proposed EPU and concludes that the proposed parameters and resultant composition and quantity of radionuclides are appropriate for the evaluation of the radioactive waste management systems.

The NRC staff further concludes that the proposed radioactive source term meets the requirements of 10 CFR Part 20, 10 CFR Part 50, Appendix I, and GDC-60. Therefore, the NRC staff finds the proposed EPU acceptable with respect to source terms.

INSERT 9 FOR SECTION 3.2 BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003

NOTE: Use Sections 2.9.2 and 2.9.3 below if the licensee's radiological consequences analyses are based on an alternative source term.

2.9.2 Radioloaical Consequences Analvses Usinq Alternative Source Terms NOTE: There are two cases that may be encountered here: (1) a licensee may be implementing an alternative source term for the first time, or (2) a licensee may have already fully implemented an alternative source term and is revising the previously approved dose analyses that use alternative source term methodologies. The second paragraph for each heading is only needed for a first-time implementation of an alternative source term (either partial or full implementations). Several accidents may have been analyzed - see corresponding SRP sections for further regulatory evaluation text (to be modified), as needed.

Requlatorv Evaluation The NRC staff reviewed the DBA radiological consequences analyses. The radiological consequences analyses reviewed are the LOCA, fuel handling accident (FHA), control rod drop accident (CRDA), and main steamline break (MSLB). The NRC staff's review for each accident analysis included (1) the sequence of events; and (2) models, assumptions, and values of parameter inputs used by the licensee for the calculation of the total effective dose equivalent (TEDE). The NRC's acceptance criteria for radiological consequences analyses using an alternative source term are based on (1) 10 CFR 50.67, insofar as it sets standards for radiological consequences of a postulated accident, and (2) GDC-19, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE, as defined in 10 CFR 50.2, for the duration of the accident. Specific review criteria are contained in SRP Section 15.0.1.

NOTE: Use the following paragraph for a first implementation of an alternative source term:

The NRC staff reviewed the implementation of alternative source terms. The NRC's acceptance criteria for implementation of alternative source terms are based on (1) 10 CFR 50.67, insofar as it sets standards for the implementation of an alternative source term in current operating nuclear power plants; (2) 10 CFR 50.49, insofar as it requires qualification of safety-related equipment, as defined in that section, including and based on integrated radiation dose during normal and accident conditions; (3) GDC-19, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE, as defined in 10 CFR 50.2, for the duration of the accident; (4) Paragraph IV.E.8 of 10 CFR Part 50, Appendix E, insofar as it requires a licensee onsite technical support center and a licensee near-site emergency operations facility from which effective direction can be given and effective control can be exercised during an emergency; and (5) plant-specific licensing commitments made in response to NUREG-0737 (Items ll.B.2, 11.6.3, II.F.l, III.D.l . I , 111.A.1.2, and lll.D.3.4). Specific review criteria are contained in SRP Sections 15.0.1.

INSERT 9 FOR SEC'I-ION 3.2 BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003

Technical Evaluation

[Insert technical evaluation. 'The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements i n the regulatory evaluation and (2) provide a clear link t o the conclusions reached by the NRC staff, as documented i n the conclusion section.]

Conclusion The NRC staff has evaluated the licensee's revised accident analyses performed in support of the proposed EPU and concludes that the licensee has adequately accounted for the effects of the proposed EPU. The NRC staff further concludes that the plant site and the dose-mitigating ESFs remain acceptable with respect to the radiological consequences of postulated DBAs since, as set forth above, the calculated total effective dose equivalent (TEDE) at the exclusion area boundary (EAB), at the low population zone (LPZ) outer boundary, and in the control room meet the exposure guideline values specified in 10 CFR 50.67 and GDC-19, as well as applicable acceptance criteria denoted in SRP Section 15.0.1. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to the radiological consequences of DBAs.

NOTE: Use the following paragraph for a first implementation of an alternative source term:

The NRC staff has reviewed the alternative source term methodology used by the licensee in evaluating the effects of the proposed EPU and concludes that changes continue to provide a sufficient margin of safety with adequate defense-in-depth to address unanticipated events and to compensate for uncertainties in accident progression, analysis assumptions, and parameter inputs. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to the implementation of an alternative source term.

INSERT 9 FOR SECTION 3.2 BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003

12.9.3 Additional Review Areas (Radiological Consequences Analyses)]

[Insert Regulatory Evaluation, Technical Evaluation, and Conclusion sections as necessary]

INSERT 9 FOR SECTION 3.2 BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003

NOTE: Use Sections 2.9.2 - 2.9.8 below if the licensee's radiological consequences analyses are not based on an alternative source term (i.e., if the analyses are based on a traditional source term (i.e., TlD- 14844) 2.9.2 Radioloqical Consequences of Control Rod Drop Accident Requlatorv Evaluation The NRC staff reviewed the analyses of the radiological consequences of a control rod drop accident (CRDA). The NRC staff's review included an examination of ( I ) the plant's response to the accident, (2) the release of fission products from the core to the environment via the turbine and condensers as a result of the accident, (3) and the calculation of radiological doses at the exclusion area boundary (EAB) and low population zone (LPZ) outer boundary, and in the control room due to the releases from the accident. The IVRC's acceptance criteria for the radiological consequences of a control rod drop accident are based on (1) GDC-19, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident, and (2) 10 CFR Part 100, insofar as it establishes requirements for assuring that radiological doses from postulated accidents will be acceptably low. Specific review criteria are contained in SRP Sections 6.4 and 15.4.9.A, and other guidance provided in Matrix 9 of RS-001.

Technical Evaluation

[Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements i n the regulatory evaluation and (2) provide a clear link t o the conclusions reached by the NRC staff, as documented in the conclusion section.]

Conclusion The NRC staff has evaluated the licensee's revised accident analyses for the radiological consequences of a control rod drop accident and concludes that the licensee has adequately accounted for the effects of the proposed EPU on these analyses. The NRC staff further concludes that the plant site and the dose-mitigatir~gESFs remain acceptable with respect to the radiological consequences of a postulated control rod drop accident since the calculated whole-body and thyroid doses at the EAB and the LPZ outer boundary are well within the exposure guideline values in 10 CFR 100.1 1. The NRC staff also concludes that the control room meets the dose requirements of GDC-19 for DBAs. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to the radiological consequences of a control rod drop accident.

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2.9.3 Radioloaical Consequences of the Failure of Small Lines Carrvinq Primarv Coolant Outside Containment Requlatorv Evaluation The NRC staff reviewed the analysis of the radiological consequences of failures outside the containment of small lines connected to the primary coolant pressure boundary (e.g., instrument lines and sample lines). The hlRC staff's review included ( I ) the identification of small lines postulated to fail and the isolation provisions for these lines; (2) the failure scenario; (3) the models and assumptions for the calculation of the radiological doses for the postulated failure; and (4) an evaluation of the primary coolant iodine activity, including the effects of a concurrent iodine spike, and the TSs for the reactor coolant iodine activity. The NRC's acceptance criteria for the radiological consequences of failures outside the containment of small lines connected to the primary coolant pressure boundary are based on (1) GDC-19, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent, to any part of the body, for the duration of the accident, and (2) GDC-55, insofar as it establishes isolation requirements for small-diameter lines connected to the primary system that form the basis of meeting 10 CFR 100.11. Specific review criteria are contained in SRP Sections 6.4 and 15.6.2, and other guidance provided in Matrix 9 of RS-001.

Technical Evaluation

[Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each o f the requirements i n the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented i n the conclusion section.]

Conclusion The NRC staff has evaluated the licensee's revised accident analyses for the radiological consequences of failures outside the containment of small lines connected to the primary coolant pressure boundary and concludes that the licensee has adequately accounted for the effects of the proposed EPU on these analyses. The NRC staff further concludes that the plant site and the dose-mitigating ESFs will remain acceptable with respect to the radiological consequences of a postulated failure outside the containment of a small line carrying reactor coolant since the calculated whole-body and thyroid doses at the EAB and the LPZ outer boundary are substantially below the exposure guideline values of 10 CFR 100.11. The NRC staff also concludes that the control room meets the dose requirements of GDC-19 for DBAs. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to the radiological consequences of failures outside the containment of small lines connected to the primary coolant pressure boundary.

INSERT 9 FOR SECTION 3.2 BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003

2.9.4 Radioloqical Consequences of Main Steamline Failure Outside Containment Requlatorv Evaluation The NRC staff reviewed the analyses of the radiological consequences of an MSLB accident outside the containment to ensure that radioactive releases due to such an event are adequately limited by the TS limit on primary coolant activity. The NRC staff's review included two cases for the reactor coolant iodine concentration: (1) an MSLB with a preaccident iodine spike and (2) an MSLB with the maximum equilibrium concentration for continued full-power operation. The NRC's acceptance criteria for the radiological consequences of an MSLB outside containment are based on (1) GDC-19, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of ,the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident, and (2) 10 CFR Part 100, insofar as it establishes requirements for assuring that radiological doses from postulated accidents will be acceptably low. Specific review criteria are contained in SRP Sections 6.4 and 15.6.4, and other guidance provided in Matrix 9 of RS-001.

Technical Evaluation

[Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]

Conclusion The NRC staff has evaluated the licensee's revised accident analyses for the radiological consequences of an MSLB outside containment and concludes that the licensee has adequately accounted for the effects of the proposed EPU on the analyses. The hlRC staff further concludes that the plant site and the dose-mitigating ESFs remain acceptable with respect to the radiological consequences of a postulated MSLB outside containment since the calculated whole-body and thyroid doses at the EAB and the LPZ outer boundary do not exceed the exposure guideline values of 10 CFR 100.11 (assuming a preaccident iodine spike) and are a small fraction of the Part 100 values for an MSLB with the primary coolant at the maximum equilibrium concentration for continued full-power operation. The NRC staff also concludes that the control room meets the dose requirements of GDC-19 for DBAs. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to a postulated failure of an MSLB outside containment.

INSERT 9 FOR SECTION 3.2 BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003

2.9.5 Radioloqical Consequences of a Desiqn-Basis Loss-of-Coolant Accident Requlatorv Evaluation The NRC staff reviewed the analyses of the radiological consequences of a design-basis LOCA. This review included a summary review of the doses from the hypothetical design-basis LOCA and a specific review of the doses from containment leakage and leakage from ESF components outside containment that contribute to the total LOCA doses. The hlRC staff's review also included (1) the contribution to the dose due to leakage from the main steam isolation valves (IVISIVs); (2) the methodology and results of calculations of the radiological consequences resulting from containment and ESF components and MSlV leakage following a hypothetical LOCA; and (3) an assessment of the containment with respect to the assumptions and the input parameters for the dose calculations. The NRC's calculations were based on pertinent information in the [Updated Safety Analysis Report or Updated Final Safety Analysis Report] and considers the NRC staff's evaluation of dose-mitigating ESFs. The NRC's acceptance criteria for the radiological consequences of a design-basis LOCA are based on (1) GDC-19, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident, and (2) 10 CFR Part 100, insofar as it establishes requirements for assuring that radiological doses from postulated accidents will be acceptably low. Specific review criteria are contained in SRP Section 6.4 and Appendices A, B, and D of SRP Section 15.6.5, and other guidance provided in Matrix 9 of RS-001.

Technical Evaluation

[Insert technical evaluation. The technical evaluation should (I) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]

Conclusion The NRC staff has evaluated the licensee's revised accident analyses for the radiological consequences of a design-basis LOCA and concludes that the licensee has adequately accounted for the effects of the proposed EPU on the analyses. The hlRC staff further concludes that the plant site and the dose-mitigating ESFs remain acceptable with respect to the radiological consequences of a design-basis LOCA since the calculated whole-body and thyroid doses at the EAB and the LPZ outer boundary do not exceed the exposure guideline values of 10 CFR 100.11 and the calculated doses in the control room meet the requirements of GDC-19. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to the radiological consequences of a design-basis LOCA.

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2.9.6 Radioloqical Consequences of Fuel Handlinq Accidents Requlatorv Evaluation The NRC staff reviewed the analyses of the radiological consequences of a postulated FHA.

The purpose of this review was to evaluate the adequacy of system design features and plant procedures provided for the mitigation of the radiologicai consequences of accidents that involve damage to spent fuel. Such accidents include the dropping of a single fuel assembly and handling tool or a heavy object onto other spent fuel assemblies. Such accidents may occur inside the containment, along the fuel transfer canal, and in the fuel building. The IVRC staff's review included (1) the sequence of events, models, and assumptions used by the licensee for the calculation of the radiological doses; (2) the adequacy of the ESFs provided for the purpose of mitigating potential accident doses; and (3) the containment ventilation system with respect to its function as a dose-mitigating ESF system, including the radiation detection system on the containment purgelvent lines for those plants that will vent or purge the containment during fuel handling operations. The MRC's acceptance criteria for the radiological consequences of FHAs are based on (1) GDC-19, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent, to any part of the body, for the duration of the accident; (2) GDC-61, insofar as it requires that systems that contain radioactivity be designed with appropriate containment, confinement, and filtering systems; and (3) 10 CFR Part 100, insofar as it establishes requirements for assuring that radiological doses from postulated accidents will be acceptably low. Specific review criteria are contained in SRP Sections 6.4 and 15.7.4, and other guidance provided in Matrix 9 of RS-001.

Technical Evaluation

['Inserttechnical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]

Conclusion The NRC staff has evaluated the licensee's revised accident analyses for the radiological consequences of FHAs and concludes that the licensee has adequately .accounted for the effects of the proposed EPU on these analyses. The hlRC staff further concludes that the plant site and the dose-mitigating ESFs remain acceptable with respect to the radiological consequences of a postulated FHA since the calculated whole-body and thyroid doses at the EAB and the LPZ outer boundary are well within the exposure guideline values of 10 CFR 100.11 and GDC-61. The NRC staff also concludes that the control room meets the dose requirements of GDC-19 for DBAs. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to the radiological consequences of FHAs.

INSERT 9 FOR SECTION 3.2 BWR TEMPLATE SAFETY EVALUATION DECEMBER 2003

2.9.7 Radioloqical Consequences of S ~ e nFuel t Cask Drop Accidents Requlatorv Evaluation The NRC staff reviewed the analyses of the radiological consequences of the release of fission products from irradiated fuel in a spent fuel cask that is postulated to drop during cask handling operations. The NRC staff's review was conducted to verify various design and operational aspects of the system. The NRC staff's review included (1) determining a need for a design-basis radiological analysis sequence of events; (2) models and assumptions used by the licensee for the calculation of the radiological doses; (3) comparing calculated doses to exposure guidelines to determine the acceptability of the EAB and LPZ outer boundary distances and to confirm the adequacy of ESFs provided for the purpose of mitigating potential doses from spent fuel cask drop accidents, includiug the effects on control room habitability; and (4) examining the relationship of the operational modes of the standby gas treatment system (SGTS) to the time sequence of the accident in order to give proper credit, in a dual containment design where the fuel building atmosphere may be exhausted through the SGTS.

The NRC's acceptance criteria for the radiological consequences of spent fuel cask drop accidents are based on (1) GDC-19, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident; (2) GDC-61, insofar as it requires that systems that contain radioactivity be designed with appropriate containment, confinement, and filtering systems; and (3) 10 CFR Part 100, insofar as it establishes requirements for assuring that radiological doses from postulated accidents will be acceptably low. Specific review criteria are contained in SRP Sections 6.4 and 15.7.5, and other guidance provided in Matrix 9 of RS-001.

Technical Evaluation

[Insert technical evaluation. The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]

Conclusion

-The NRC staff has evaluated the licensee's revised accident analyses for the radiological consequences of a spent fuel cask drop accident and concludes that the licensee has adequately accounted for the effects of the proposed EPU on these analyses. The NRC staff further concludes that the plant site and the dose-mitigating ESFs remain acceptable with respect to the radiological consequences of a postulated spent fuel cask drop accident since the calculated whole-body and thyroid doses at the EAB and the LPZ outer boundary are well within the exposure guideline values of 10 CFR 100.11 and GDC-61. The NRC staff also concludes that the control room meets the dose requirements of GDC-19 for DBAs. Therefore, the t4RC staff finds the licensee's proposed EPU acceptable with respect to spent fuel cask drop accidents.

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[2.9.8 Additional Review Areas (Source Terms and Radiological Consequences Analyses)]

[Insert Regulatory Evaluation, Technical Evaluation, and Conclusion sections as necessary]

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