1CAN051401, Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One, Unit 1 at 60 Years Arkansas Nuclear One Unit 1

Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One, Unit 1 at 60 Years Arkansas Nuclear One Unit 1

ML14126A816
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/06/2014
From:	Pyle S Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
1CAN051401 ANP-3281NP, Rev 1
Download: ML14126A816 (30)

Text

Attachment 1 to this letter contains proprietary information - Attachment 1 is withheld from public disclosure per 10 CFR 2.390.

1CAN051401 May 6, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852

SUBJECT:

Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One, Unit 1 at 60 Years Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51

REFERENCES:

1. Entergy letter dated January 31, 2000, License Renewal Application, (1CAN010003) (ML003679667)

2. BAW-2248A, Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, April 2000 (ML003708443)

3. Entergy letter dated August 24, 2000, License Renewal Application RAIs, (1CAN080003) (ML003746995)

4. NRC letter dated April 12, 2001, Arkansas Nuclear One, Unit 1 - License Renewal Safety Evaluation Report, (ML011030091)

5. Duke Energy letter dated February 20, 2012, License Renewal Commitment to Submit a Time Limiting Aging Analysis for the Reactor Vessel Internals to the NRC for Review, (ML12053A332)

Dear Sir or Madam:

By letter dated January 31, 2000 (Reference 1), Entergy Operations, Inc. (Entergy) submitted a License Renewal Application (LRA) for Arkansas Nuclear One, Unit 1 (ANO-1). As noted in the LRA, BAW-2248A provides a description of the reactor vessel internals for ANO-1. BAW-2248A was developed on a generic basis for several Babcock & Wilcox units, including ANO-1, to demonstrate that the aging effects for the reactor vessel internals are adequately managed for the period of extended operation.

Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4704 Stephenie L. Pyle Manager, Regulatory Assurance Arkansas Nuclear One

1CAN051401 Page 2 of 3 to this letter contains proprietary information - Attachment 1 is withheld from public disclosure per 10 CFR 2.390.

As a result of the NRC review of Reference 2, several Renewal Applicant Action Items were identified. The ANO-1 specific response to Renewal Applicant Action Item #12 states A plant-specific analysis will be performed to demonstrate that under LOCA and seismic loading, the internals have adequate ductility to absorb local strain at the regions of maximum stress intensity and that irradiation accumulated at the expiration of the renewal license will not affect deformation limits. Data will be developed to demonstrate that the internals will meet the deformation limits at the expiration of the renewed license.

Reference 3 transmitted responses to NRC Requests for Additional Information (RAIs) pertaining to the reactor vessel and reactor vessel internals. The response to RAI 4.1-1 states The TLAA reported in BAW-2248A regarding ductility of stainless steel and deformation limits will be evaluated by Entergy Operations when sufficient embrittlement data is collected through the BWOG and EPRI MRP programs. Once the embrittlement data is available, the TLAA reported in BAW-2248A will be updated. The TLAA will be resolved using 10 CFR 54.21(c)(1)(iii), which is consistent with the approach taken by Duke Power for the Oconee Units that was approved by the NRC (NUREG-1723, page 4-24). At present, ANO plans to complete the evaluation prior to the end of the current term of operation.

The NRC Safety Evaluation (Reference 4) stated that the staff found that the applicants responses to the Renewal Applicant Action Items resolve the action items from BAW-2248.

Specifically, for Item #12, the staff stated in part that the analysis will be performed as part of the applicants reactor vessel internals aging management program (RVIAMP).

ANO-1 enters the period of extended operation on May 20, 2014.

provides the ANO-1 specific time-limited aging analysis for the loss of ductility of the reactor vessel internals. This analysis demonstrates that the internals will meet the deformation limits at the expiration of the renewal license and fulfills the commitment made in Reference 1. The remaining portion of the RVIAMP will be provided under a separate cover letter.

This analysis is considered to be proprietary to AREVA, Inc. AREVA, Inc. requests that the proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390.

AREVA, Inc. has provided Entergy with authorization to provide the proprietary information. An affidavit by the information owner, AREVA, Inc., supporting the request for non-disclosure is provided in Attachment 2. Therefore, Entergy requests that Attachment 1 of this submittal be withheld from public disclosure in accordance with 10 CFR 2.390. Attachment 3 provides a non-proprietary version of the analysis.

This request and analysis are similar to those provided by Duke Energy via Reference 5.

1CAN051401 Page 3 of 3 to this letter contains proprietary information - Attachment 1 is withheld from public disclosure per 10 CFR 2.390.

This letter contains no new regulatory commitments.

If you have any questions or require additional information, please contact me.

Sincerely, Original signed by Stephenie L. Pyle SLP/rwc

Attachment:

1. Time-Limited Aging Analysis Regarding Reactor Vessel Internal Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years - PROPRIETARY

2. Affidavit

3. Time-Limited Aging Analysis Regarding Reactor Vessel Internal Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years - NON-PROPRIETARY cc:

Mr. Marc L. Dapas Regional Administrator U. S. Nuclear Regulatory Commission, Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Peter Bamford MS O-8B3 One White Flint North 11555 Rockville Pike Rockville, MD 20852

to 1CAN051401 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years PROPRIETARY

to 1CAN051401 Affidavit

to 1CAN051401 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years NON-PROPRIETARY

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report March 2014 AREVA Inc.

(c) 2014 AREVA Inc.

Controlled Document

Copyright © 2014 AREVA Inc.

All Rights Reserved Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page i Nature of Changes Item Section(s) or Page(s)

Description and Justification Rev. 0 All Initial Issue Rev. 1 All Added assumptions and results sections, renumbered other sections accordingly Nomenclature New acronyms added 1.0 New quotations from and discussions regarding ANO-1 LR documentation 5.0 Added RV internals at the end of the last sentence 7.0 Added References 3, 4, and 6, updated Reference 5 Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page ii Contents Page

1.0 INTRODUCTION

............................................................................................... 1-1

2.0 BACKGROUND

................................................................................................. 2-1 3.0 ASSUMPTIONS................................................................................................ 3-1 4.0 INPUTS............................................................................................................. 4-1 5.0 ANALYSIS......................................................................................................... 5-1 6.0 RESULTS.......................................................................................................... 6-1

7.0 REFERENCES

.................................................................................................. 7-1 Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page iii List of Figures Figure 5-1 Effect of Radiation on Uniform Elongation of Annealed Type 304 Stainless Steel Irradiated at 290°C (554°F) and Tested at Various Temperatures......................................................................................... 5-2 Figure 5-2 [

]........................................................................... 5-3 Figure 5-3 Change in Uniform Elongation as a Function of Neutron Dose for Solution-Annealed Type 304, 304L, and 347 Stainless Steels at Elevated Temperatures (270-380°C)...................................................... 5-4 Figure 5-4 Effect of Strain Rate and Temperature on the Uniform Elongation of Type 304 Stainless Steel........................................................................ 5-5 Figure 5-5 Effect of Strain Rate and Temperature on the 0.2 Percent Yield Stress of Type 304 Stainless Steel......................................................... 5-5 Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page iv Nomenclature Acronym Definition AMP Aging Management Program AMR Aging Management Review ANO-1 Arkansas Nuclear One Unit 1 ASTM American Society of Testing and Materials B&W B[&]WOG Babcock and Wilcox Babcock and Wilcox Owners Group (now Pressurized Water Reactor Owners Group, or PWROG)

CFR CLB Code of Federal Regulations Current Licensing Basis EFPY Effective Full-Power Years EPRI Electric Power Research Institute LOCA Loss Of Coolant Accident LRA License Renewal Application LWR Light Water Reactor MeV Million Electron Volts NRC Nuclear Regulatory Commission RAl Request for Additional Information RCS Reactor Coolant System RV RVIAMP Reactor Vessel Reactor Vessel Internals Aging Management Programa SC Structures and Components SER TLAA Safety Evaluation Report Time-Limited Aging Analysis UFSAR Updated Final Safety Analysis Report a RVIAMP is used in the direct quotes from the NRC to indicate Reactor Vessel Internals (RVI) Aging Management Program (AMP). In the AREVA text, other forms such as RV internals or reactor vessel internals are used to avoid confusion with Reactor Vessel Integrity (RVI), as commonly used in other AREVA reports.

Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 1-1

1.0 INTRODUCTION

Entergy Operations submitted a License Renewal Application (LRA) (Reference 1) in January 2000 for Arkansas Nuclear One Unit 1 (ANO-1), which provided the technical information as required by 10 CFR 54. The application intended to provide sufficient information for the Nuclear Regulatory Commission (NRC) to complete its technical reviews. BAW-2248A (Reference 2) was developed on a generic basis for several Babcock and Wilcox (B&W) units, including ANO-1, to demonstrate that the aging effects for the reactor vessel (RV) internals within the scope of Reference 2 are adequately managed for the period of extended operation.

As detailed in the LRA, the NRC issued several renewal application action items as the result of their review of BAW-2248A. Renewal applicant action item #12 reads as follows:

Plant-specific analysis is required to demonstrate that, under loss-of-coolant-accident (LOCA) and seismic loading, the internals have adequate ductility to absorb local strain at the regions of maximum stress intensity and that irradiation accumulated at the expiration of the renewal license will not adversely affect deformation limits. The RVIAMP must develop data to demonstrate that the internals will meet the deformation limits at that expiration of the renewal license.

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 1-2 As a result of this requirement, Entergy provided the following response within the LRA (see Section 2.3.1.6 and Table 2.3-5 of Reference 1):

A plant-specific analysis will be performed to demonstrate that under LOCA and seismic loading, the internals have adequate ductility to absorb local strain at the regions of maximum stress intensity and that irradiation accumulated at the expiration of the renewal license will not affect deformation limits. Data will be developed to demonstrate that the internals will meet the deformation limits at the expiration of the renewed license.

This requirement is captured as a time-limited aging analysis (TLAA) in Table 4.1-1 of the LRA (Reference 1) and referenced in the reactor vessel internals aging management program in Appendix B of Reference 1. In response to RAI 4.1-1 provided by the NRC staff on April 25, 2000 regarding the timing and means of how this TLAA will be addressed (Reference 3), ANO-1 provided the following response on August 24, 2000 (Reference 4):

The TLAA reported in BAW-2248A regarding ductility of stainless steel and deformation limits will be evaluated by Entergy Operations when sufficient embrittlement data is collected through the BWOG and EPRI MRP programs.

Once the embrittlement data is available, the TLAA reported in BAW-2248A will be updated. The TLAA will be resolved using 10CFR54.21(c)(1)(iii), which is consistent with the approach taken by Duke Power for the Oconee Units that was approved by the NRC (NUREG-1723, page 4-24). At present, ANO plans to complete the evaluation prior to the end of the current term of operation.

The NRCs safety evaluation report (SER) for the ANO-1 LRA (Reference 5) provided the following in Section 3.3.2.4.3:

Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 1-3 On the basis of the review described above, the staff finds that the applicant has demonstrated that the effects of aging associated with the reactor vessel internals will be adequately managed so that there is reasonable assurance that the intended function will be maintained with the CLB for the period of extended operation.

The NRCs safety evaluation report (SER) for the ANO-1 LRA (Reference 6) provided the following in Section 4.1.3:

The NRC staff concludes that the applicant has provided a list of acceptable TLAAs as defined in 10 CFR 54.3, and that no 10 CFR 50.12 exemptions have been granted on the basis of a TLAA as defined in 10 CFR 54.3.

This document provides the plant-specific evaluation of ductility for the RV internals at the expiration of the renewed license, which is 54 Effective Full-Power Years (EFPY) using projected fluence values for ANO-1, as required per LRA and SER.

Information considered proprietary to AREVA is marked in square brackets, [ ]

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 2-1

2.0 BACKGROUND

In 2010, Appendix E of the 1970 RV internals topical report was updated by AREVA through a contract with the Electric Power Research Institute (EPRI) for 60 years on a generic basis for the B&W units and submitted, for information, to the NRC.

(Reference 7) This update identified the locations of the maximum stress intensity where a loss of ductility because of neutron irradiation would be detrimental (from Appendix E of the 1970 RV internals topical report) as the core barrel flanges.

However, upon more detailed examination of the wording and stress intensity values presented in the 1970 RV internals topical report, the location of highest stress intensity occurs at the core support shield bottom flange.

The excerpts describing such locations from the 1970 RV internals topical report, Section 3.2.3.2 are below:

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 2-2

[

] Therefore, the bottom core support shield flange will be evaluated within this document as the region of maximum stress intensity, as required by renewal applicant action item #12 from BAW-2248A.

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 3-1 3.0 ASSUMPTIONS There are no assumptions for this document.

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 4-1 4.0 INPUTS This section identifies and provides inputs to the ANO-1-specific evaluation of ductility for the RV internals at 54 EFPY. The first required input is a projected fluence value specifically generated for ANO-1 at 54 EFPY. The methodology used to determine the neutron fluence was based on AREVAs NRC approved fluence analysis methodology, described in topical report BAW-2241P-A (References 8, 9, and 10). The fluence methodology is consistent with the guidance of Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.

As described in Section 2.0 of this document, the bottom of the core support shield (i.e.,

at the lower flange) is the location of interest and the projected 54 EFPY fluence for this location at ANO-1 is [

] Using the light water reactor (LWR) conversion factor of 15 dpa = 1 x 1022 n/cm2 (E>1.0 MeV) (Reference 11), this converts to [

] In addition, it is noted that the fluence at the top of the core support shield (i.e., at the upper flange) would be less than at the bottom of the core support shield because of the increased distance from the core.

The second required input is the material used for the manufacturing of the core support shield top and bottom flanges. As detailed in Reference 2, the core support shield flanges are fabricated from American Society of Testing and Materials (ASTM) A 473-63 Type 304 austenitic stainless steel.

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 5-1 5.0 ANALYSIS The projected 54 EFPY fluence is [

] for the core support shield lower flange, which has been shown to be the region of maximum stress intensity for the RV internals; the fluence at the top of the core support shield would be less than this value because of the increased distance from the core.

Figure 5-1 (Figure E-3 of the 1970 RV internals topical report) depicts that for a fluence of [

] Note that Figure 5-2 (Figure 3-12 in Reference 7) provides recent irradiated Type 304 test data to compare to the curves in Figure 5-1. The test data validates the conservatism of the curves in Figure 5-1. This slight decrease in uniform elongation at this level of fluence is confirmed in Figure 5-3 (Figure 13(c) of Reference 11).

In addition, the uniform elongation of unirradiated solution annealed Type 304 stainless steel at 600°F is seen to only decrease slightly with increasing strain rate as shown in Figure 5-4 (Figure 5 of Reference 12). However, even at the highest tested strain rates, at 600°F, the uniform elongation is above the 20 percent uniform elongation of irradiated material credited for 40 years in Appendix E of the 1970 RV internals topical report and the 8.6 percent allowable strain specified in Appendix A of the 1970 RV internals topical report. It is also observed that yield strength increases with increasing strain rate at 600°F as shown in Figure 5-5 (Figure 3 of Reference 12). In addition to having sufficient ductility at 60 years relative to the allowables of the 1970 RV internals topical report, the upper and lower core support shield flanges will have greater resistance to plastic deformation at increased strain rates.

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 5-2 Therefore, the conclusions from Appendix E to the 1970 RV internals topical report concerning the acceptable ductility and deformation limits for a 40-year lifetime remain valid for a 60-year lifetime for the ANO-1 RV internals.

Figure 5-1 Effect of Radiation on Uniform Elongation of Annealed Type 304 Stainless Steel Irradiated at 290°C (554°F) and Tested at Various Temperatures Note: This is Figure E-3 in Appendix E of the 1970 RV internals topical report.

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 5-3 Figure 5-2 [

]

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 5-4 Figure 5-3 Change in Uniform Elongation as a Function of Neutron Dose for Solution-Annealed Type 304, 304L, and 347 Stainless Steels at Elevated Temperatures (270-380°C)

Note: This is Figure 13(c) in Reference 11.

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 5-5 Figure 5-4 Effect of Strain Rate and Temperature on the Uniform Elongation of Type 304 Stainless Steel Note: This is Figure 5 in Reference 12.

Figure 5-5 Effect of Strain Rate and Temperature on the 0.2 Percent Yield Stress of Type 304 Stainless Steel Note: This is Figure 3 in Reference 12.

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 6-1 6.0 RESULTS By NRC submittal, Entergy Operations committed to providing a plant-specific analysis to demonstrate that under LOCA and seismic loading, the internals have adequate ductility to absorb local strain at the regions of maximum stress intensity and that irradiation accumulated at the expiration of the renewed license will not affect deformation limits for ANO-1 prior to the end of the current term of operation. A projected fluence value at 54 EFPY was developed for ANO-1. Based on material data included in a 1970 RV internals topical report and newer data, the conclusions from Appendix E to the 1970 RV internals topical report concerning the acceptable ductility and deformation limits for a 40-year lifetime remain valid for a 60-year lifetime for the ANO-1 RV internals.

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 7-1

7.0 REFERENCES

1. Arkansas Nuclear One - Unit 1 Docket No. 313, License No. DPR-51, License Renewal Application, January 31, 2000. NRC Accession Number ML003679667.

2. BAW-2248A, Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, April 2000. NRC Accession Number ML003708443.

3. Request for Additional Information for the Review of the Arkansas Nuclear One, Unit 1, License Renewal Application, April 25, 2000, NRC Accession Number ML003707431.

4. Arkansas Nuclear One - Unit 1, Docket No. 50-313, License No. DPR-51, License Renewal Application RAIs (TAC No. MA8064), August 24, 2000, NRC Accession Number ML003746995.

5. NUREG-1743, Safety Evaluation Report Related to License Renewal of Arkansas Nuclear One, Unit 1, Chapter 3, May 2001, NRC Accession Number ML011640177.

6. NUREG-1743, Safety Evaluation Report Related to License Renewal of Arkansas Nuclear One, Unit 1, Chapter 4-End, May 2001, NRC Accession Number ML011640217.

7. Information in Support of the EPRI Materials Reliability Program (MRP):

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0) Review, October 29, 2010, NRC Accession Number ML103090248.

8. Letter from James F. Malley to NRC Document Control Desk, Submittal of BAW-2241P, Revision 2, Fluence and Uncertainty Methodologies, June 2, 2003, NRC Accession Number ML031550365.

Controlled Document

AREVA Inc.

ANP-3281NP Revision 1 Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years Licensing Report Page 7-2

9. BAW-2241NP-A, Revision 2, Fluence and Uncertainty Methodologies, April 30, 2006, NRC Accession Number ML073310660.

10. Letter from Ronnie L. Gardner to NRC Document Control Desk, Publication of Revision 1 of Appendix G to BAW-2241(P), Revision 2, Fluence and Uncertainty Methodologies, November 20, 2007, NRC Accession Number ML073310655.

11. NUREG/CR-7027, Degradation of LWR Core Internal Materials Due to Neutron Irradiation, December 2010, NRC Accession Number ML102790482.

12. High Strain Rate Tensile Properties of AISI Type 304 Stainless Steel, J.

M. Steichen, Journal of Engineering Materials and Technology, July 1973.

Controlled Document