ML082630509

2008/09/19 - Oyster Creek, Safety Evaluation Report Related to License Renewal, Supplement 1

ML082630509
Person / Time
Site:	Oyster Creek
Issue date:	09/19/2008
From:	Baty M C NRC/OGC
To:	NRC/OCM
SECY RAS
References
50-219-LR, RAS H-72
Download: ML082630509 (25)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE COMMISSION

In the Matter of )

)

AMERGEN ENERGY COMPANY, LLC ) Docket No. 50-219-LR

)

(Oyster Creek Nuclear Generating Station) )

CERTIFICATE OF SERVICE I hereby certify that copies of the "SAFETY EVALUATION REPORT RELATED TO THE LICENSE RENEWAL OF OYSTER CREEK NUCLEAR GENERATING STATION, SUPPLEMENT 1" have been served on the following by electronic mail with copies by deposit in the NRC's internal mail system or, as indicated by an asterisk, by electronic mail, with copies by U.S. mail, first class, this 19 th day of September, 2008.

E. Roy Hawkens, Chair Administrative Judge

Atomic Safety and Licensing Board

Mail Stop: T-3F23

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

E-mail: ERH@nrc.gov

Anthony J. Baratta

Administrative Judge

Atomic Safety and Licensing Board

Mail Stop: T-3F23

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

E-mail: AJB5@nrc.gov

Paul B. Abramson

Administrative Judge

Atomic Safety and Licensing Board

Mail Stop: T-3F23

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

E-mail: PBA@nrc.gov

Office of the Secretary

ATTN: Docketing and Service

Mail Stop: O-16G4

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

E-mail: HEARINGDOCKET@nrc.gov

Office of Commission Appellate

Adjudication

Mail Stop: O-16G4

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

E-mail: OCAAMail@nrc.gov

Emily Krause

Law Clerk

Atomic Safety and Licensing Board

Mail Stop: T-3F23

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

E-mail: EIK1@nrc.gov

Suzanne Leta Liou*

New Jersey Public Interest Research Group

11 N. Willow St.

Trenton, NJ 08608

E-mail: sliou@environmentnewjersey.org

Donald Silverman, Esq.*

Alex S. Polonsky, Esq.

Kathryn M. Sutton, Esq.

Raphael P. Kuyler, Esq.

Morgan, Lewis & Bockius LLP

1111 Pennsylvania Ave., N.W.

Washington, DC 20004

E-mail: dsilverman@morganlewis.com apolonsky@morganlewis.com ksutton@morganlewis.com rkuyler@morganlewis.com

Paul Gunter, Director*

Kevin Kamps

Reactor Watchdog Project

Nuclear Information

And Resource Service

6930 Carroll Avenue Suite 340

Takoma Park, MD 20912

E-mail: paul@beyondnuclear.org kevin@beyondnuclear.orq

J. Bradley Fewell, Esq.*

Exelon Corporation

4300 Warrenville Road

Warrenville, IL 60555

E-mail: bradley.fewell@exeloncorp.com

Richard Webster, Esq.*

Julia LeMense, Esq.*

Rutgers Environmental Law Clinic

123 Washington Street

Newark, NJ 07102-5695

Email: rwebster@easternenvironmental.org j lemense@easternenvironmental.org

/RA/ ______________________________

Mary C. Baty

Counsel for the NRC Staff

Safety Evaluation Report Related to the License Renewal of Oyster Creek Nuclear Generating Station Supplement 1 Docket No. 50-219

AmerGen Energy Company, LLC

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation

September 2008 THIS PAGE INTENTIONALLY LEFT BLANK.

iii ABSTRACT This document is a supplemental safety eval uation report (SSER) for the license renewal application (LRA) for Oyster Creek Nuclear Gener ating Station (OCGS) as filed by AmerGen Energy Company, LLC. (AmerGen or the applicant

). By letter dated July 22, 2005, AmerGen submitted its application to the United States (US) Nuclear Regulatory Commission (NRC) for

renewal of the OCGS operating license for an additional 20 years. The NRC staff issued a final

safety evaluation report (SER) in two volumes, dated March, 2007(ML071290023 and

ML071310246), which summarizes the results of its safety review of the renewal application for

compliance with the requirements of Title 10, Part 54, of the Code of Federal Regulations , (10 CFR Part 54), "Requirements for Renewal of Operating Licenses for Nuclear Power Plants."

This document only lists the changes to the March 2007 SER.

This SSER documents clarifications provided by the applicant to drywell shell aging

management program commitments. The applicant provi ded these clarifications as a result of the Atomic Safety and Licensing Board hearing, which was held September 24-25, 2007 in

Toms River, NJ. This SSER also documents the staff's evaluation of the applicant's response to

an April 2008 request for additional information (RAI) regarding the use of Green's function in

calculating fatigue cumulative usage factors (CUF) for OCGS components.

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TABLE OF CONTENTS Abstract........................................................................................................................................iii Abbreviations...............................................................................................................................vi

1.0 Introduction

and General Discussion.................................................................................1-1

2.0 Structures

Systems and Components...............................................................................2-1

3.0 Aging

Management Results...............................................................................................3-1 4.0 Time Limited Aging Analysis..............................................................................................

4-1 4.3.4 Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping.......................................................4-1 4.3.4.1 Summary of Technical Information in the Application..........................4-1 4.3.4.2 Staff Evaluation....................................................................................4-1 4.3.4.3 UFSAR Supplement.............................................................................4-5 4.3.4.4 Conclusion............................................................................................4-5

5.0 Review

by the Advisory Committee on Reactor Safeguards.............................................5-1 6.0 Conclusion.........................................................................................................................6-1 Appendices

Appendix A: Commitments fo r License R enewal.........................................................A-1

vi ABBREVIATIONS

ACRS Advisory Committee on Reactor Safeguards AmerGen AmerGen Energy Company, LLC AMP aging management program ASLB Atomic Safety and Licensing Board ASME American Society of Mechanical Engineers

CFR Code of Federal Regulations CUF cumulative usage factor

DO dissolved oxygen

EAF environmentally assisted fatigue

HWC hydrogen water chemistry

F en fatigue life correction factor

LAS low-alloy steel LRA license renewal application

NRC Nuclear Regulatory Commission NWC normal water chemistry

OCGS Oyster Creek Generating Station

RAI request for additional information RO reactor recirculation outlet nozzle

SC structures and components SIA Structural Integrity Associates SER safety evaluation report SSER supplemental safety evaluation report

3-D three dimensional

US United States UFSAR updated final safety analysis report 1-1 SECTION 1 INTRODUCTION AND GENERAL DISCUSSION

1.1 Introduction

This document is a supplemental safety eval uation report (SSER) for the license renewal application (LRA) for Oyster Creek Nuclear Gener ating Station (OCGS) as filed by AmerGen Energy Company, LLC. (AmerGen or the applicant

). By letter dated July 22, 2005, AmerGen submitted its application to the United States (US) Nuclear Regulatory Commission (NRC) for

renewal of the OCGS operating license for an additional 20 years. The NRC staff (the staff)

issued a final safety evaluation report (SER) in two volumes, dated March, 2007 (ADAMS

Accession Nos. ML071290023 and ML071310246), which summarizes the results of its safety

review of the renewal application for compliance with the requirements of Title 10, Part 54, of

the Code of Federal Regulations , (10 CFR Part 54), "Requirements for Renewal of Operating Licenses for Nuclear Power Plants." This SSER supplements portions of Sections 1, 4, 6, and

Appendix A of the March 2007 SER.

In a letter dated January 14, 2008 (ML080160540), the applicant provided clarifications to

commitments related to the aging management program for the OCGS drywell shell, associated

with AmerGen's renewal application. This letter was a result of the Atomic Safety and Licensing

Board hearing, which was held September 24-25, 2007 in Toms River, NJ. Clarifications made

to these commitments are included in Appendix A. Commitment number 27-2 changed the

enhancement from "Refueling Outages" to "Prior to filling the reactor cavity with water."

Commitment number 27-18 expanded the "required thickness values" to "Code-specified safety

factors" for "refueling load cases and post-accident load case." Commitment 27-22 specifies

that the applicant will verify that the sand bed drain lines are clear from obstruction every other

refueling outage.

Additionally, on April 29, 2008, the staff issued a request for additional information (RAI) to the

applicant requesting they address the staff's concerns with regard to the use of Green's function

in calculating fatigue cumulative usage factors (CUF) for OCGS components. The Staff's

concerns with licensee use of Green's function are discussed in detail in Draft NRC Regulatory Issue Summary 2008-XX, "Fatigue Analysis of Nuclear Power Plant Components," dated April

11, 2008 (ML080950235). Section 4 of this SSER contains the staff's evaluation of the

applicant's response to that RAI.

1.7

SUMMARY

OF PROPOSED LICENSE CONDITIONS The staff does not have any changes or update to this section of the original SER (ML071290023 and ML071310246).

2-1 SECTION 2 STRUCTURES SYSTEMS AND COMPONENTS

The staff does not have any changes or update to this section of the original SER (ML071290023 and ML071310246).

THIS PAGE INTENTIONALLY LEFT BLANK 3-1 SECTION 3 AGING MANAGEMENT REVIEW RESULTS The staff does not have any changes or update to this section of the original SER (ML071290023 and ML071310246).

THIS PAGE INTENTIONALLY LEFT BLANK 4-1 SECTION 4 TIME LIMITED AGING ANALYSIS

4.3.4 Effects

of Reactor Coolant Envir onment on Fatigue Life of Components and Piping (Generic Safety Issue 190) 4.3.4.1 Summary of Technical Information in the Application The staff does not have any changes or update to this section of the original SER (ML071290023 and ML071310246).

4.3.4.2 Staff Evaluation Subsequent to the issuance of NUREG-1875 in April 2007, the staff issued a request for

additional information (RAI) on April 29, 2008 (ADAMS Accession No. ML081080077), to

address concerns identified with the use of Green's function in calculating fatigue cumulative

usage factors (CUF) of Oyster Creek Generating Station (OCGS) components.

As a matter of background, the Green's function approach involves performing a detailed stress

analysis of a component to calculate its response to a step change in temperature. This detailed

analysis is used to establish an influence functi on, which is subsequently used to calculate the stresses caused by actual plant temperature tr ansients. In implementing this approach, OCGS used a simplified input for applying the Green's function in which only one value of stress was

used for the evaluation of the actual plant transients. Detailed stress analysis, however, requires

consideration of six stress components, as discussed in the American Society of Mechanical

Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code),Section III, Subarticle NB-

3200. Thus, the concern is whether the simplified implementation of the Green's function

method provides acceptable and conservative results. AmerGen used the simplified input for

applying the Green's function to perform fatigue calculations at OCGS for the reactor

recirculation outlet (RO) nozzles.

The staff's RAI asked the applicant to perform a confirmatory detailed fatigue analysis (herein

referred to as "confirmatory analysis") of the RO nozzles in accordance with the methodology of

ASME Code,Section III, Subarticle NB-3200, which retains all six stress components, to

determine whether the analysis that employed the simplified Green's function methodology (herein referred to as "original analysis"), which uses a single stress term, provided acceptable

results. The staff's RAI also asked that the applicant confirm that the RO nozzles were the only

components where the Green's function was used to evaluate the fatigue CUF for license

renewal.

4-2 In its response to the staff's RAI, dated May 1, 2008 (ML081270386), the applicant stated that it used the Green's function methodology in the OCGS license renewal only to evaluate the

fatigue CUF for the RO nozzles. Therefore, in the case of OCGS license renewal, the concern

related to the simplified Green's function methodology involves only the RO nozzles.

The applicant performed the requested confirmatory analysis. The following four items show the

applicant's description of the differences between the original analysis and the confirmatory

analysis, stated in its response to the staff's RAI, and the staff's view of the applicant's

statements.

(1) The applicant stated that Green's function was not used in the confirmatory analysis and that all six components of stress were extracted from a finite element analysis of all

transients and then used in calculating the fatigue usage factor in accordance with

Subparagraph NB-3216.2 in Section III of the ASME Code. The applicant also stated:

"Calculated stresses are comparable and CUF is lower."

The staff audited the calculations underlying the confirmatory analysis and determined

that the confirmatory analysis calculations were performed in accordance with the ASME

Code,Section III, Subarticle NB-3200. The staff found that the stress analysis calculated

for all six stress components were extracted from the stress results of the ANSYS model

at the nozzle limiting location which were subsequently used in the fatigue calculation of

the nozzle. The staff also found that calculated stresses were comparable to the original

analysis and the fatigue evaluation resulted in a lower CUF.

(2) The applicant stated that in the original analysis for the OCGS RO nozzles, stresses were conservatively extracted on the stainless steel cladding surface and were

evaluated using the carbon steel fatigue curve, which provided very conservative fatigue

usage results. The applicant also stated that in the confirmatory analysis, "the nozzle

cladding was neglected for the fatigue calculation, as permitted in NB-3122.3 of Section

III of the ASME Code, and the base metal was evaluated for stresses and fatigue

usage." The applicant stated that this was "consistent with the approach used in

NUREG/CR-6260 for several component evaluations."

The staff notes that ASME Code Section III, Subparagraph NB-3122.3, directs that for

stresses and fatigue evaluation purposes the presence of the cladding should be

considered in the analyses and both materials, cladding and base metal, shall meet the

code stresses and fatigue requirements. However, if the integrally bonded cladding is

not considered to add to the component's structural integrity, as the code states "is 10%

or less of the total thickness of the component", Subparagraph NB-3122.3 allows the

analyst to exclude the cladding from the stress and fatigue evaluation and evaluate only

the stresses and fatigue on the base metal.

The staff confirmed during the audit that in the confirmatory analysis the cladding was

included in the ANSYS finite element model and stresses were extracted on the base

metal surface adjacent to cladding (base metal stresses), which were then used to

evaluate the base metal for fatigue usage. This is permitted by the ASME Code.

Therefore, the staff finds this approach acceptable.

4-3 (3) The applicant stated that in performing the confirmatory fatigue evaluation for the emergency condenser transients, an input of hold time was assumed in the analysis, between the initial downward shock in temperature and the subsequent warm-up. The

original analysis did not include this hold time based on plant-specific transient

evaluation. The applicant stated that it made this change to conservatively ensure that

the peak stress is captured after the downward shock, and to address all possible

scenarios of event severity for future plant operation.

The staff finds this method acceptable as it addresses steady state periods and captures peak stresses. Furthermore, the staff's audit confirmed that this change resulted in a

higher stress for those transients in the confirmatory analysis, however, as the applicant

stated in its response, and the Staff confirmed, this increase in stress was an

insignificant contributor to fatigue usage compared to the decrease in fatigue usage

described by item 2 above.

(4) For the confirmatory environmentally assisted fatigue (EAF) analysis, the applicant stated that it used detailed environmental fatigue factor (F en) multipliers that were determined for each load pair based on maximum transient temperatures with assumed

low strain rates, resulting in a maximized strain rate term for each of these F en multipliers. In the original EAF analysis, detailed F en multipliers were determined for each load pair based upon both maximum transient temperatures and calculated strain rates.

This change was made to conservatively determine the strain rates and the resultant

environmental fatigue multipliers for the confirmatory analysis. This change resulted in

an increase in the overall F en multiplier for the OCGS RO nozzles in the confirmatory analysis.

During its audit, the staff noted that the F en multipliers, as calculated in the original analysis using strain rates calculated for each load pair, were in accordance with

NUREG/CR-6583, "Effects of LWR Coolant Environments on Fatigue Design Curves of

Carbon and Low-Alloy Steels", issued February 1998, and in the confirmatory analysis

the NUREG/CR-6583 bounding strain rate of 0.001 percent per second (%/s) was used

for the F en multipliers in all load pairs. Therefore, the staff finds the strain rate values used in both analyses acceptable. In addition, the staff noted that using the maximum

transient temperatures resulted in conservative F en multiplier values, and is, therefore, acceptable.

In its response to the staff's RAI the applicant submitted tabulated results of both the original

analysis and the confirmatory analysis. The tables show that the confirmatory analysis resulted in an EAF CUF of 0.1366 for the nozzle corner location versus the original analysis EAF CUF of

0.9781 at the same location. The detailed confirmatory analysis found the original analysis

results conservative by a factor of approximately seven.

The staff audited the applicant's calculations in the confirmatory and original analyses, reviewing plant data and analysis records. The staff's audit review showed that both analyses

used the same ANSYS axisymmetric finite element model. In addition, the staff found that the

applicant had verified that its contractor, Structural Integrity Associates (SIA), who performed

both the original and confirmatory analyses, had also validated that both analyses utilized the

same finite element model. The staff also confirmed that the tabulated results shown in the RAI 4-4 response tables for the original analysis and the confirmatory analysis were for the inside corner of the nozzle at the blend radius on the cladding surface and on the base metal surface

adjacent to cladding respectively.

The staff reviewed the methodology to calculate the F en multipliers in both the original and confirmatory analyses and found that both used the same low-alloy steel (LAS) equation shown

in NUREG/CR-6583 and developed detailed F en multipliers for each load pair. Therefore, the staff finds the methodology of deriving the F en multipliers acceptable.

The staff also reviewed the four inputs (sulfur content, temperature, strain rate and dissolved

oxygen (DO)) used to develop the F en multipliers as specified in NUREG/CR-6583 for LAS. Both analyses used the NUREG/CR-6583 bounding sulfur content of 0.015 weight percent. In the

original analysis F en values were determined based upon maximum transient temperatures and calculated strain rates. In the confirmatory analysis F en values were determined based upon maximum transient temperatures and an assumed slow strain rate equal to the NUREG/CR-

6583 bounding strain rate of 0.001%/s. The staff finds this acceptable because it produces a

higher, thus conservative, overall F en multiplier in the confirmatory analysis.

The staff found that both analyses accounted for the DO water chemistry effect for hydrogen

water chemistry (HWC) as well as the normal wa ter chemistry (NWC) employed by the applicant prior to the implementation of HWC. The staff reviewed available plant data for both HWC and

NWC. F en values were based on 59 percent HWC and 41 percent NWC. The staff's Audit of the available chemistry data showed that this is a conservative representation of Oyster Creek's water chemistry. The staff also reviewed plant data for DO and verified that the appropriate

values were used to calculate the F en multipliers in accordance with NUREG/CR-6583 for LAS and is, therefore, acceptable. In addition, the applicant stated that water chemistry (HWC and

noble metals chemistry) and water chemistry monito ring will ensure that DO levels remain below the value used to calculate the F en values.

Based on its review, the staff finds that the applicant adequately accounted for the water

chemistry effects in the evaluation of EAF.

The staff's review of the applicant's calculations verified that the confirmatory analysis resulted

in a CUF of 0.0207 and a EAF CUF of 0.1366 for the limiting nozzle corner location compared to the original analysis CUF of 0.1832 and EAF CUF of 0.9781 at the same location. The original

analysis conservatively used stresses from the stainless steel cladding surface, which were

then used for the fatigue evaluation of the LAS base metal. The applicant stated that the

assumptions in the original analysis resulted in a conservative estimate of the fatigue usage. In

the confirmatory analysis, the stainless steel cladding was excluded (see above under item 2)

for the fatigue calculation, as permitted in Subparagraph NB-3122.3 in Section III of the ASME

Code, and the base metal stresses were used for the fatigue evaluation.

The staff agrees that the applicant's fatigue evaluation which uses the base metal stresses, and

not the cladding stresses, to perform the fatigue analysis of the base metal is in accordance

with the ASME Code Section III.

From the staff's review of the calculations and from discussions with the applicant's contractor

SIA during the audit, the staff determined that the primary difference between the original and 4-5 confirmatory analyses, in addition to the use of all six stress components in the confirmatory analysis, was that in the original analysis, stresses were extracted at the cladding surface and it

was assumed that those were the stresses at the base metal surface beneath the cladding, while in the confirmatory analysis, stresses were extracted at the base metal surface itself.

Thus, because the ASME Code does not require the conservative assumption they made in the

original analysis, and because the stresses at the cladding surface are greater than those at the

base metal surface (at the interface with the cladding), the original analysis was more

conservative than it needed to be. That factor contributed the most to the decreased fatigue

CUF in the confirmatory analysis.

The staff's audit confirmed that the original analysis results are conservative by a factor of

approximately seven when compared to the detailed confirmatory analysis results. Based on the staff review, the original analysis is acceptable.

4.3.4.3 UFSAR Supplement The staff does not have any changes or update to this section of the original SER (ML071290023 and ML071310246).

4.3.4.4 Conclusion Based on its review of selected analysis records, the staff finds that the applicant's derivation of

F en values is in accordance with NUREG/CR-6583 and that the applicant has adequately accounted for reactor water chemistry effects in calculating the EAF usage for the RO nozzles.

In addition, the staff determined that the confirmatory analysis is in accordance with the rules

and requirements of ASME Code,Section III, Subarticle NB-3200, and is, therefore, acceptable.

Furthermore, the detailed confirmatory analysis found the original analysis results conservative

and, therefore, acceptable.

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5-1 SECTION 5 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

The staff has provided the Advisory Committee on Reactor Safeguards with a copy of this Supplemental Safety Evaluation Report.

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6-1 SECTION 6 CONCLUSION On the basis of its review of the LRA, the staff concludes that the requirements of

10CFR§54.29(a) have been met. The staff concludes that the confirmatory analysis performed

by the applicant, to address concerns with the use of Green's function in calculating fatigue

cumulative usage factors (CUF) for OCGS components, is acceptable.

A-1 APPENDIX A COMMITMENTS FOR LICENSE RENEWAL OF OCGS

During the review of the Oyster Creek Generating Station (OCGS) license renewal application (LRA) by the staff of the United States (US) Nuclear Regulatory Commission (NRC) (the staff),

AmerGen Energy Company, LLC. (the applic ant) made commitments related to aging management programs (AMPs) to manage the agi ng effects of structures and components (SCs) both prior to and during the period of extended operation. The following table lists revised

commitments 27-2, 27-18 as well as a new commitment 27-22.

A-2 COMMITMENT NUMBER ITEM NUMBER AND COMMITMENT UFSAR SUPPLEMENT LOCATION (LRA APP. A)

ENHANCEMENT OR IMPLEMENTATION SCHEDULE SOURCE 27) ASME Section XI, Subsection

IWE (2) A strippable coating will be applied to the reactor cavity liner to prevent water intrusion into the gap

between the drywell shield wall and the drywell shell

during periods when the reactor cavity is flooded.

(18) AmerGen will perform a 3-D finite element structural analysis of the primary containment drywell shell

using modern methods and current drywell shell

thickness data to better quantify the margin that

exists above the required minimum for buckling. The

analysis will include sensitivity studies to determine

the degree to which uncertainties in the size of

thinned areas affect Code margins. If the analysis

determines that the drywell shell does not meet

Code-specified safety factors (i.e., 2.0 for the

refueling load case and 1.67 for the post-accident

load case), the NRC will be notified in accordance

with 10 CFR 50 requirements.

(22) Verify that the sand bed drain lines are clear from obstruction.

A.1.27 Prior to filling the

reactor cavity with

water.

Prior to the period of

extended operation.

Every other refueling

outage. Section B.1.27

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