GNRO-2016/00034, License Renewal Documents to Support Review of Application for Renewal

License Renewal Documents to Support Review of Application for Renewal

ML16285A303
Person / Time
Site:	Grand Gulf
Issue date:	10/10/2016
From:	Fallacara V Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML16286A154	List: GNRO-2016/00034, License Renewal Documents to Support Review of Application for Renewal
References
GNRO-2016/00034
Download: ML16285A303 (210)

Text

Ala

=eEntergy Entergy Operations, Inc.

P. O. Box 756 Port Gibson, MS 39150 Vincent Fallacara Site Vice-President Grand Gulf Nuclear Station Tel. (601) 437-2129 GNRO-2016/00034 October 10, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Grand Gulf Nuclear Station License Renewal Documents To Support Review Of Application For Renewal Of License: Proprietary and Non-Proprietary Versions of Report Number MPM-814779, Revision 3, Neutron Transport Analysis For Grand Gulf Nuclear Station, and Engineering Report GGNS-NE-15-00003, Revision 0, Grand Gulf Nuclear Station Fluence Effect on RPV Internal Components At EPU Operating Conditions Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 License No. NPF-29

REFERENCES:

1. Entergy Letter GNRO-2014/00076, "Response to Request for Additional Information (RAI) Set 51" dated November 6, 2014

2. U.S. NRC Letter, "Request for Additional Information for the Review of Grand Gulf Nuclear Station, License Renewal Application, Set 52" dated April 6, 2015 (GNRI-2015/00020)

3. Entergy Letter GNRO-2015/00034, "Response to Request for Additional Information (RAI) Set 52" dated May 20, 2015

4. U.S. NRC Letter, "Summary of Telephone Conference Call Held On June 18, 2015, Between The U.S. NRC And Entergy Concerning Request For Additional Information Responses, Pertaining To The Grand Gulf Nuclear Station License Renewal Application (TAC NO. ME7493)

5. Entergy Letter GNRO-2015/00048, "Response to License Renewal Amendment Request for Additional Information (RAI) Set 47, Question 4.2.1-2c (5) (b)", dated July 29, 2015

6. Entergy Letter GNRO-2015/00055, "Responses to Request for Additional Information (RAI) Set 52, RAls 3.0.3-1-FWS-2a and 3.0.3-2b", dated August 19, 2015

7. U.S. NRC Letter, "Requests for Additional Information for the Review of the Grand Gulf Nuclear Station License Renewal Application (TAC NO.

ME7493) - SET 53", dated October 28, 2015 (GNRI-2015/00125)

8. Entergy Letter GNRO-2015/00079, "Responses to Request for Additional Information (RAI) Set 53", dated November 23, 2015

9. Entergy Letter GNRO-2016/00018, "Grand Gulf Nuclear Station License Renewal Documents To Support Review Of Application For Renewal Of License", dated April 4, 2016.

GNRO-20 16/00034 Page 2 of 3

Dear Sir or Madam:

Entergy Operations, Inc. is providing, in the Attachments, Proprietary and Non-Proprietary versions of Report Number MPM-814779, Revision 3, "Neutron Transport Analysis For Grand Gulf Nuclear Station," and Proprietary and Non-Proprietary versions of Engineering Report GGNS-NE-15-00003, Revision 0, "Grand Gulf Nuclear Station Fluence Effect on RPV Internal Components At EPU Operating Conditions" to support your review of Grand Gulf's License Renewal Application. Affidavits for the Proprietary versions are also attached.

This letter contains no new commitments.

If you have any questions or require additional information, please contact James Nadeau at 601-437-2103.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 10th day of October, 2016.

VF/ras Attachments:

1. Report Number MPM-814779, Revision 3, "Neutron Transport Analysis For Grand Gulf Nuclear Station", Proprietary version

2. Report Number MPM-814779, Revision 3, "Neutron Transport Analysis For Grand Gulf Nuclear Station", Non-Proprietary (redacted) version

3. NEDC-33669P, Revision 1, April 2015, Proprietary Version, "Grand Gulf Nuclear Station Fluence Effect on RPV Internal Components At EPU Operating Conditions", (Attachment 1 to Engineering Report GGNS-NE-15-00003, Revision 0)

4. NEDO-33669, Revision 1, April 2015, Non-Proprietary Version, "Grand Gulf Nuclear Station Fluence Effect on RPV Internal Components At EPU Operating Conditions",

(Non-Proprietary Version of Attachment 1 to Engineering Report GGNS-NE-15-00003, Revision 0)

5. Affidavit for Proprietary Version of Report Number MPM-814779, Revision 3

6. Affidavit for Proprietary Version of NEDC-33669P, Revision 1, April 2015, (Attachment 1 to Engineering Report GGNS-NE-15-00003, Revision 0)

GNRO-2016/00034 Page 3 of 3 cc: with Attachments U.S. Nuclear Regulatory Commission ATTN: Mr. Manny Sayoc, NRR/DLR Project Manager Office of License Renewal Mail Stop 0-11 F1 Washington, DC 20555 cc: without Attachments U.S. Nuclear Regulatory Commission ATTN: Mr. Kriss Kennedy Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 U.S. Nuclear Regulatory Commission ATTN: Mr. J. Kim, NRR/DORL Mail Stop OWFN/8 G14 11555 Rockville Pike Rockville, MD 20852-2378 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150

Attachment 2 to GNRO-2016/00034 Report Number MPM-814779, Revision 3, "Neutron Transport Analysis For Grand Gulf Nuclear Station",

Non-Proprietary (redacted) version

NON-PRUPRlETARY VER~'10N Report Number MPM-814779 Revision 3 Neutron Transport Analysis For Grand Gulf Nuclear Station I

t;u;it£(liteTUI

& TESTING, LLC February, 2015

ivon-rroprtetary MPM Technical Report Neutron Transport Analysis for Grand Gulf Nuclear Station O

MPM Report Number MPM-814779 Revision 3 Final Report, February, 2015 MP Machinery and Testing, LLC 2161 Sandy Drive State College, PA 16803 Redacted information is identified by blank space enclosed with double brackets [[ ]].

MP Machinery and Testing, LLC *2161 Sandy Drive, State College, Pennsylvania 16803* USA 814-234-8860

tvon-rroprtetary

[[

]]

DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THIS DOCUMENT WAS PREPARED BY MP MACHINERY AND TESTING, LLC AS AN ACCOUNT OF WORK SPONSORED OR COSPONSORED BY ENTERGY OPERATION, INC.. NEITHER MPM, ANY COSPONSOR, NOR ANY PERSON ACTING ON BEHALF OF ANY OF THEM:

(A) MAKES ANY WARRANTY OR REPRESENTATION WHATSOEVER, EXPRESS OR IMPLIED, (I) WITH RESPECT TO THE USE OF ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT, INCLUDING MERCHANTABILITY AND FITNESS FOR A PARTICULAR PURPOSE, OR (II) THAT SUCH USE DOES NOT INFRINGE ON OR INTERFERE WITH PRIVATELY OWNED RIGHTS, INCLUDING ANY PARTY'S INTELLECTUAL PROPERTY, OR (III) THAT THIS DOCUMENT IS SUITABLE TO ANY PARTICULAR USER'S CIRCUMSTANCE; OR (B) ASSUMES RESPONSIBILITY FOR ANY DAMAGES OR OTHER LIABILITY WHATSOEVER (INCLUDING ANY CONSEQUENTIAL DAMAGES, EVEN IF MPM OR ANY MPM REPRESENTATIVE HAS BEEN ADVISED OF THE POSSIBILITY OF SUCH -DAMAGES) RESULTING FROM YOUR SELECTION OR USE OF THIS DOCUMENT OR ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR 'ITEM DISCLOSED IN THIS DOCUMENT.

ivon-rroprtetary Nuclear Quality Assurance Certification

[[

]] All work has been performed under the MPM Nuclear Quality Assurance Program.

M. P. Manahan, Sr.

President 2/20/2015 Date J. Nemet QA Manager 2/20/2015 Date 11

tvon-rroprtetary EXECUTIVE

SUMMARY

This work was undertaken to calculate the best estimate neutron fluence, and its uncertainty, to the Grand Gulf Nuclear Station (GGNS) reactor pressure vessel, (RPV) core shroud and top guide horizontal and vertical welds, as well as to several beltline vessel nozzles.

Figures ES-l and ES-2 show the shroud and top guide weld designations, and Figure ES-3 shows the weld seam and plate locations for the reactor vessel. The fluence calculations were carried out using a three dimensional (3D) neutron transport model for each fuel cycle starting from cycle 1 through cycle 21. The 3D neutron transport calculations were benchmarked on a plant-specific basis by comparing calculated results against previously performed core region 2D synthesis data as well as by calculation of the calculated-to-measured (C/M) ratios for GGNS dosimetry. In addition, a comprehensive benchmarking report of MPM methods has been prepared under separate cover.

The neutron transport calculational procedures and dosimetry analysis methods meet standards specified by the NRC and ASTM as appropriate. In particular, the transport analysis meets the requirements of Regulatory Guide 1.190 (RG 1.190). Since RG 1.190 is focused on 2D synthesis methods, it is strictly applicable to analyses in the active fuel region. Nevertheless, the guidance provided in RG 1.190 was followed to the extent practical for modeling work in the regions above and below the active fuel region. The 3D neutron transport calculations were used to determine detailed fluence profiles at the end of cycle 21 (28.088 EFPY), and projected to exposures of35 EFPY and 54 EFPY. [[

]]

Summary ofShroud and Top Guide Fluence Results

[[

]]

111

tvon-rroprietary

[[

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Summary of Vessel, Vessel Internals, and Cycle 1 Dosimetry Results The transport calculations were also performed to evaluate fluence for the surveillance capsule and for the reactor vessel. Comparisons with dosimetry measurements at the GGNS surveillance capsule location at the end of cycle 1 were made and excellent agreement was found. [[

]]

The vessel has nozzle penetrations at several locations, and neutron exposure at the nozzles is of concern for neutron damage analysis. Four sets of nozzles were evaluated. For nozzles below the core, the maximum fluence point occurs at the top of the nozzle. The reverse IV

non-rroprtetary is true for nozzles above the core. [[

]]

In addition to the fluence analysis for the RPV, calculation offluences at nineteen component locations was completed in support of the RPV internals mechanical evaluation. The calculations focused on the determination of peak values for the structures and components specified to ensure conservatism in the calculations. In the case of the shroud, the peak on the ID surface of the shroud is reported. For plates, such as the core plate, the peak anywhere on the plane defining the plate has been calculated. For components with discrete angular positions, such as the shroud support legs, the peak reported is the peak over the 360 degrees at the elevation of the legs. For the components with specified axial ranges, such as the core plate bolts and the top guide bolts, the peak was determined as well as the average over the axial extent of the bolts at the peak azimuthal location.

Benchmarking and Uncertainty Analysis Fluence values for the capsule, vessel, and shroud in the beltline region (except for the very top and bottom of the core) are estimated to have uncertainties of [[ ]] These uncertainties are within the value of +/- 20% specified by RG 1.190. Moreover, the 3D calculations have been benchmarked [[

]]

Regulatory Guide 1.190 requires that the overall fluence calculation bias and uncertainty must be determined by an appropriate combination of the analytic uncertainty analysis results and the results of the uncertainty analysis based on the comparisons to the operating reactor and simulator benchmark measurements. The regulatory guide states that this combination may be a weighted average that accounts for the reliability of the individual estimates. The regulatory guide goes on to state that if the analytical uncertainty at the 1 sigma level is greater than 30%,

the methodology of the regulatory guide is not applicable and the application will be reviewed on an individual basis. [[

]]

v

tvon-rroprietary Table ES-l Maximum Fluence to GGNS Shroud and Top Guide Horizontal Welds.

[[

]]

VI

non-rraprtetary Table ES-2 Maximum Fluence to GGNS Shroud and Top Guide Vertical Welds.

[[

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VB

rcon-rropnetary Table ES-3 GGNS Maximum Calculated Vessel Fluence and Fluence with dpa Attenuation.

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>51 VB TOP CUIOE fLAHCE i I Figure ES-l Weld Designations for GGNS Top Guide.

SHROUO TOP cua: nANG[

H3 ~'----"""":.lIl--_ ......-----~.a-.--ri o o VlJ 0 CENTRAl. UPPER CYlINDER l-1.00 ....SWATCH t.fAXIWt.f CENtRAL WIODLE CY\.INO(R Vt5 V16 H5 t - - - - - - - -......-------~

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Figure ES-2 Weld Designations for GGNS Shroud.

IX

non-rroprtetary

[[

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Figure ES-3 GGNS Vessel Rollout Drawing Showing Beltline Weld Seam and Plate Locations. The Increase in the Axial Extent of the Beltline Requires that Radiation Damage Effects for Shell I, 2 and 3 Materials be Included in the P-T Curve Analysis.

x

tvon-rroprtetary CONTENTS 1 INTRODUCTION 1-1 1.1 Section 1 References 1-2 2 NEUTRON TRANSPORT MODEL DESCRIPTION 2-1 2.1 Source Representation 2-1 2.2 Neutron Transport Model 2-3 2.3 Section 2 References 2-7 3 BENCHMARK ANALYSES 3-1 3.1 Compliance with RG 1.190 3-1 3.2 Summary of MPM Methods 3-5 3.3 Benchmarking MPM Methodology 3-6 3.4 Conclusions 3-9 3.5 Section 3 References 3-9 4 SHROUD AND TOP GUIDE FLUENCE RESULTS 4-1 4.1 Shroud Weld Fluence Results 4-1 4.2 Comparison with Past Results 4-3 4.3 Section4 References 4-4 5 VESSEL AND CAPSULE FLUENCE RESULTS 5-1 5.1 Cycle 1 Dosimetry Analysis 5-1 5.2 Pressure Vessel Analysis 5-2 5.3 Comparison of RPV Peak Fluence with Past Results 5-5 5.4 RPV Internals Analysis 5-5 5.5 Section 5 References 5-6 6 UNCERTAINTY ANALYSIS 6-1 6.1 Uncertainty Assumptions 6-1 6.2 20/30 Uncertainty Evaluation over Active Fuel Length 6-5 Xl

non-rroprtetary 6.3 3D Model Uncertainty Evaluation for Shroud, Top Guide Weld, and Nozzle Locations 6-6 6.4 Uncertainty Conclusions 6-8 6.4 Section6 References 6-8 7

SUMMARY

AND CONCLUSiONS 7-1 8 NOMENCLATURE 8-1 A SHROUD/TOP GUIDE WELD FLUENCE RESULTS AT THE END OF CYCLE 21 (28.088 EFPY EXPOSURE)

• A-1 B SHROUD/TOP GUIDE WELD FLUENCE RESUL1S AFTER 35 EFPY EXPOSURE B-1 C SHROUD/TOP GUIDE WELD FLUENCE RESUL 1S AFTER 54 EFPY EXPOSURE C-1 xu

tvon-rroprtetary LIST OF FIGURES Figure 1-1 Weld Designations for GGNS Top Guide 1-3 Figure 1-2 Weld Designations for GGNS Shroud 1-3 Figure 1-3 GGNS Vessel Rollout Drawing Showing Beltline Weld Seam and Plate Locations. The Increase in the Axial Extent of the Beltline Requires that Radiation Damage Effects for Shell 1, 2 and 3 Materials be Included in the P-T Curve Analysis 1-4 Figure 2-1 GGNS R-8 Geometry used in the TORT Calculations 2-15 Figure 2-2 Diagram Showing Locations of Fuel Nodes versus Axial Height Relative to BAF 2-16 Figure 4-1 Fast Flux (E>1.0 MeV) at the Shroud IR to Weld H4 for Cycles 1 through 21. 4-8 Figure 4-2 Fast Fluence (E>1.0 MeV) to Shroud Weld H4 at the End of Cycle 21 (28.088 EFPY) 4-9 Figure 4-3 Fast Fluence (E>I.0 MeV) to Shroud Weld H4 at 35 EFPY 4-10 Figure 4-4 Fast Fluence (E> 1.0 MeV) to Shroud Weld H4 at 54 EFPY 4-11 Figure 4-5 Fast Fluence (E>1.0 MeV) to Shroud Vertical Welds V13-V16 at the End of Cycle 21 (28.088 EFPY) 4-12 Figure 4-6 Fast Fluence (E>1.0 MeV) to Shroud Vertical Welds V13-V16 at 35 EFPY 4-13 Figure 4-7 Fast Fluence (E>1.0 MeV) to Shroud Vertical Welds V13-V16 at 54 EFPY 4-14 Figure 5-1 Azimuthal Variation of Maximum Vessel Fluence (E > 1.0 MeV) at the End of Cycle 21 (28.088 EFPY) 5-42 Figure 5-2 Axial Variation of Peak Fluence (E > 1.0 MeV) in the Vessel at the End of Cycle 21 (28.088 EFPY) 5-43 Figure 5-3 GGNS Vessel Roll Out Drawing Showing Beltline Weld Seam and Plate Locations. The Increase in the Axial Extent of the Beltline will Eventually Require that Radiation Damage Effects for Shell 3 Materials be Included in the PT Curve Analysis 5-44 Figure 5-4 Axial Variation of Peak Fluence (E > 1.0 MeV) in the Vessel at 54 EFPY 5-45 Xlll

ivon-rroprtetary LIST OF TABLES Table 2-1Neutron Energy Group Structure used in the GGNS Transport Calculations - 47 Groups ' 2-9 Table 2-2 GGNS Reactor Component Radial Dimensions 2-10 Table 2-3GGNS Reactor Component Azimuthal Locations 2-13 Table 2-4GGNS Reactor Component Axial Dimensions 2-14 Tab'le 3-1 Summary of Regulatory Guide 1.190 Positions on Fluence Calculation Methods 3-11 Table 3-2 Summary of Regulatory Guide 1.190 Positions on Fluence Measurement Methods 3-13 Table 3-3 Summary of Regulatory Guide 1.190 Positions on Fluence and Uncertainty Reporting 3-14 Table 3-4 Comparison of Calculated and Measured Results for PCA 3-15 Table 3-5 Comparison of Calculated and Benchmark Results for the Reactor Vessel Calculated at Reactor Axial Peak. 3-16 Table 3-6 Tabulation of Dosimetry Results for Operating BWR Plants 3-17 Table 4-1 Locations for Fluence Evaluation in the Shroud and Top Guide Horizontal Welds 4-5 Table 4~2 Locations for Fluence Evaluation in the Shroud and Top Guide Vertical Welds 4-5 Table 4-3 Estimated Maximum Fluence to GGNS Shroud and Top Guide Horizontal Welds 4-6 Table 4-4 Estimated Maximum Fluence to GGNS Shroud and Top Guide Vertical Welds 4-7 Table 5-1Calculated Fe(n,p) Reaction Rates at the Cycle 1 Dosimetry Location 5-7 Table 5-2 Flux Spectrum at Dosimetry Location :1....*****.********** 5-8 Table 5-3 Grand Gulf Cycle 1 Power History 5-9 Table 5-4Nuclear Parameters Used in the Evaluation of Neutron Sensors 5-10 Table 5-5Tabulation of Dosimetry Results 5-10 Table 5-6 Azimuthal Variation of Maximum Fluence at the Vessel Wetted Surface 5-11 Table 5-7 Axial Variation of Maximum Fluence at the Vessel Wetted Surface 5-18 Table 5-8 Relative Exposure through Vessel at Maximum Fluence Point at the End of Cycle 21 (28.088 EFPY) ~ 5-27 XIV

ivon-rroprtetary Table 5-9 Calculated Maximum Vessel Exposure (Shell 2) at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures 5-28 Table 5-10 Azimuthal Location Ranges of Plates in Vessel Shell 1, Shell2, and Shell 3 5-29 Table 5-11 Calculated Maximum Vessel Shell 1 Exposure at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures 5-30 Table 5-12 Calculated Maximum Vessel Shell 3 Exposure at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures 5-31 Table 5-13 Azimuthal Locations of Vertical Welds in Vessel Shell 1, Shell 2 and Shell 3 5-32 Table 5-14 Calculated Maximum Vessel Shell 1 Weld Exposures at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures 5-32 Table 5-15 Calculated Maximum Vessel Shell 2 Weld Exposures at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures 5-33 Table 5-16 Calculated Maximum Vessel Shell 3 Weld Exposures at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures 5-33 Table 5-17Calculated Maximum Vessel Circumferential Weld Exposures at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures 5-34 Table 5-18 GGNS Calculated Vessel Fluence and Fluence Determined using dpa Attenuation 5-34 Table 5-19 GGNS RPV Nozzle Locations 5-35 Table 5-20 GGNS RPV Nozzle Maximum Fluence Values 5-35 Table 5-21GGNS RPV Nozzle Region Maximum Fluence Values in the First Quadrant at the Axial Locations Specified in Reference [5-9] 5-36 Table 5-22GGNS RPV Nozzle Region Fluence Values at the Axial and Azimuthal Locations Specified in Reference [5-9] 5-36 Table 5-23 GGNS RPV Nozzle N6 Maximum Fluence Values With Margin Term Applied to Account for Above Core Fluence Uncertainty 5-37 Table 5-24 Fluence Locations Specified by GEH for RPV Internals Mechanical Evaluation 5-38 Table 5-25 Fluence Evaluation at the End of Cycle 21 (28.088 EFPY) for use in RPV Internals Mechanical Evaluation 5-39 Table 5-26 Fluence Evaluation at the 35 EFPY for use in RPV Internals Mechanical Evaluation 5-40 Table 5-27 Fluence Evaluation at the 54 EFPY for use in RPV Internals Mechanical Evaluation 5-41 Table 6-1 Grand Gulf Shroud, Capsule, and Vessel Active Fuel Region TORT Calculational Fluence Uncertainty 6-10 Table 6-2 Grand Gulf Shroud, Capsule, and Vessel Active Fuel Region 2D SYnthesis Calculational Fluence Uncertainty 6-11 Table 6-3 Estimated Maximum Uncertainty for Shroud and Top Guide Horizontal Welds 6-12 xv

tvon-rroprtetary Table 6-4 Estimated Maximum Uncertainty for Shroud and Top Guide Vertical Welds........ 6-12 Table 6-5 Estimated Maximum Weld IR Fluence Uncertainty for Shroud and Top Guide Vertical Welds 6-13 Table 6-6 Estimated Maximum Uncertainty forVessel Nozzles N6 and NI2 6-13

,/

XVI

ivon-rropnetary LIST OF APPENDIX A TABLES SHROUD/TOP GUIDE WELD FlUENCE RESULT5 AT THE END OF CYCLE 21 (28.088 EFPY EXPOSURE)

Appendix Table A- 1 Fast Fluence at Locations in the Top Guide for Weld HI vs.

Azimuth A-2 Appendix Table A- 2 Fast Fluence at Locations in the Top Guide for Weld H2 vs.

Azimuth A-3 Appendix Table A- 3 Fast Fluence at Locations in the Shroud for Weld H3 vs. Azimuth A-4 Appendix Table A- 4 Fast Fluence at Locations in the Shroud for Weld H4 vs. Azimuth A-5 Appendix Table A- 5 Fast Fluence at Locations in the Shroud for Weld H5 vs. Azimuth" A-6 Appendix Table A- 6 Fast Fluence at Locations in the Shroud for Weld H6A vs.

Azimuth" A-7 Appendix Table A- 7 Fast Fluence at Locations in the Shroud for Weld H6B vs.

Azimuth A-8 Appendix Table A- 8 Fast Fluence at Locations in the Shroud for Weld H7 vs. Azimuth A-9 Appendix Table A- 9 Fast Fluence at Locations in the Top Guide for Welds VI and V3 vs. Height above BAF A-I0 Appendix Table A- 10 Fast Fluence at Locations in the Top Guide for Welds V2 and V4 vs. Height above BAF A-I0 Appendix Table A- 11 Fast Fluence at Locations in the Top Guide for Weld V5 and V6 vs. Height above BAF A-II Appendix Table A- 12 Fast Fluence at Locations in the Top Guide for Welds V7 and V8 vs. Radial Location A-12 Appendix Table A- 13 Fast Fluence at Locations in the Top Guide for Welds V9 and VII vs. Height above BAF A-13 Appendix Table A- 14 Fast Fluence at Locations in the Top Guide for Welds VI0 and V12 vs. Height above BAF A-13 Appendix Table A- 15 Fast Fluence at Locations in the Shroud for Weld V13 and V14 vs. Height above BAF A-14 Appendix Table A- 16 Fast Fluence at Locations in the Shroud for Weld V15 and V16 vs. Height above BAF A-15 Appendix Table A- 17 Fast Fluence at Locations in the Shroud for Welds V17 and V18 vs. Height above BAF A-16 XVll

lVon-rroprietary Appendix Table A- 18 Fast Fluence at Locations in the Shroud for Welds V19 and V21 vs. Height above BAF A-I7 Appendix Table A- 19 Fast Fluence at Locations in the Shroud for Welds V20 and V22 vs. Height above BAF A-I7 Appendix Table A- 20 Fast Fluence at Locations in the Shroud for Welds V23 and V24 vs. Height above BAF A-I8 XVlll

tvon-rroprtetary LIST OF APPENDIX B TABLES SHROUD/TOP GUIDE WELD FLUENCE RESUL15 AFTER 35 EFPY EXPOSURE Appendix Table B- 1 Fast Fluence at Locations in the Top Guide for Weld HI vs.

Azimuth B-2 Appendix Table B- 2 Fast Fluence at Locations in the Top Guide for Weld H2 vs.

Azimuth B-3 Appendix Table B- 3 Fast Fluence at Locations in the Shroud for Weld H3 vs. Azimuth B-4 Appendix Table B- 4 Fast Fluence at Locations in the Shroud for Weld H4 vs. Azimuth B-5 Appendix Table B- 5 Fast Fluence at Locations in the Shroud for Weld H5 vs. Azimuth" B-6 Appendix Table B- 6 Fast Fluence at Locations in the Shroud for Weld H6A vs.

Azimuth" B-7 Appendix Table B- 7 Fast Fluence at Locations in the Shroud for Weld H6B vs. Azimuth B-8 Appendix Table B- 8 Fast Fluence at Locations in the Shroud for Weld H7 vs. Azimuth B-9 Appendix Table B- 9 Fast Fluence at Locations in the Top Guide for Welds VI and V3 vs. Height above BAF B-l 0 Appendix Table B- 10 Fast Fluence at Locations in the Top Guide for Welds V2 and V4 vs. Height above BAF B-I0 Appendix Table B- 11 Fast Fluence at Locations in the Top Guide for Weld V5 and V6 vs. Height above BAF B-ll Appendix Table B- 12 Fast Fluence at Locations in the Top Guide for Welds V7 and V8 vs. Radial Location B-12 Appendix Table B- 13 Fast Fluence at Locations in the Top Guide for Welds V9 and VII vs. Height above BAF B-13 Appendix Table B- 14 Fast Fluence at Locations in the Top Guide for Welds VI0 and V 12 vs. Height above BAF B-13 Appendix Table B- 15 Fast Fluence at Locations in the Shroud for Weld V13 and V14 vs.

Height above BAF B-14 Appendix Table B- 16 Fast Fluence at Locations in the Shroud for Weld V15 and V16 vs.

Height above BAF B-15 Appendix Table B- 17 Fast Fluence at Locations in the Shroud for Welds V17 and V18 vs. Height above BAF B-16 Appendix Table B- 18 Fast Fluence at Locations in the Shroud for Welds V19 and V21 vs. Height above BAF B-l 7 XIX

tvon-rropnetary Appendix Table B- 19 Fast Fluence at Locations in the Shroud for Welds V20 and V22 vs. Height above BAF B-17 Appendix Table B- 20 Fast Fluence at Locations in the Shroud for Welds V23 and V24 vs. Height above BAF B-18 xx

tvon-rropnetary LIST OF APPENDIX C TABLES SHROUD/TOP GUIDE WELD FLUENCE RESULTS AFTER 54 EFPY EXPOSURE Appendix Table C- 1 Fast Fluence at Locations in the Top Guide for Weld HI vs.

Azimuth C-2 Appendix Table C- 2 Fast Fluence at Locations in the Top Guide for Weld H2 vs.

Azimuth C-3 Appendix Table C- 3 Fast Fluence at Locations in the Shroud for Weld H3 vs. Azimuth C-4 Appendix Table C- 4 Fast Fluence at Locations in the Shroud for Weld H4 vs. Azimuth C-5 Appendix Table C- 5 Fast Fluence at Locations in the Shroud for Weld H5 vs. Azimuth" C-6 Appendix Table C- 6 Fast Fluence at Locations in the Shroud for Weld H6A vs.

Azimuth" C-7 Appendix Table C- 7 Fast Fluence at Locations in the Shroud for Weld H6B vs. Azimuth C-8 Appendix Table C- 8 Fast Fluence at Locations in the Shroud for Weld H7 vs. Azimuth C-9 Appendix Table C- 9 Fast Fluence at Locations in the Top Guide for Welds VI and V3 vs. Height above BAF C-l 0 Appendix Table C- 10 Fast Fluence at Locations in the Top Guide for Welds V2 and V4 vs. Height above BAF C-I0 Appendix Table C- 11 Fast Fluence at Locations in the Top Guide for Weld V5 and V6 vs. Height above BAF C-l1 Appendix Table C- 12 Fast Fluence at Locations in the Top Guide for Welds V7 and V8 vs. Radial Location C-12 Appendix Table C- 13 Fast Fluence at Locations in the Top Guide for Welds V9 and VII vs. Height above BAF C-13 Appendix Table C- 14 Fast Fluence at Locations in the Top Guide for Welds VI0 and V12 vs. Height above BAF C-13 Appendix Table C- 15 Fast Fluence at Locations in the Shroud for Weld V13 and V14 vs.

Height above BAF C-14 Appendix Table C- 16 Fast Fluence at Locations in the Shroud for Weld V15 and V16 vs.

Height above BAF C-l5 Appendix Table C- 17 Fast Fluence at Locations in the Shroud for Welds V17 and V18 vs. Height above BAF C-16 Appendix Table C- 18 Fast Fluence at Locations in the Shroud for Welds V19 and V21 vs. Height above BAF C-17 XXI

tvon-rroprtetary Appendix Table C- 19 Fast Fluence at Locations in the Shroud for Welds V20 and V22 vs. Height above BAF C-17 Appendix Table C- 20 Fast Fluence at Locations in the Shroud for Welds V23 and V24 vs. Height above BAF C-18 XXll

tvon-rropnetary 1

INTRODUCTION This work was undertaken to calculate the best estimate neutron tluence, and its uncertainty, to the Grand Gulf Nuclear Station (GGNS) top guide, reactor pressure vessel (RPV),

vessel nozzles, and core shroud horizontal and vertical welds. The neutron exposure at the top guide, shroud, and, in particular, the shroud welds is an important concern for many Boiling Water Reactors (BWRs) since cracks have been observed in several plants. Figures 1-1 and 1-2 show the GGNS top guide and shroud weld designations, and Figure 1-3 shows the GGNS weld seam and plate locations for the reactor vessel. These designations are used throughout the report to identify the vertical and horizontal welds of interest. It is important to note that, with the exception of welds V7 and V8, all of the shroud and top guide welds are located within the cylindrical portion of these components. [[

]]

The neutron tluence calculations were carried out using a three dimensional (3D)TORT model for each fuel cycle through the end of cycle 21. At the time of the calculations, the plant had not completed cycle 20, and therefore the fuels data has been extrapolated to the estimated end of cycle 20. It will be necessary to update the cycle 20 transport analysis sometime in the future to represent the as-burned cycle. [[

]] The neutron transport models are fully described in Section 2. The calculational procedures meet standards specified by the Nuclear Regulatory Commission (NRC) and American Society for Testing and Materials (ASTM) as appropriate. In particular, the beltline analysis meets the requirements of Regulatory Guide 1.190 (RG 1.190) [1-2]. An important requirement of RG 1.190 is methodology benchmarking. In addition to the plant-specific benchmarking for GGNS, other MP Machinery and Testing, LLC (MPM) benchmark work is summarized in Section 3. A full discussion of the benchmarking ofMPM methods is given in Reference [1-3].

1-1

Introduction Non-Proprietary The shroud fluence results are presented in Section4 and the Appendices. The vessel fluence calculations are given in Section5. RG 1.190 requires that a detailed uncertainty analysis be performed to identify each source of uncertainty and the impact on the overall accuracy of the calculation. The uncertainty analysis results are reported in Section6. Summary and conclusions are given in Section7.

1.1 Section 1 References

[1-1] [[

]]

[1-2] Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U. S. Nuclear Regulatory Commission, March 2001.

[1-3] [[

]]

1-2

Introduction lVon-jDroprieta~

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O' Figure 1-2 Weld Designations for GGNS Shroud.

1-3

Introduction Non-Proprietary

[[

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Figure 1-3GGNS Vessel Rollout Drawing Showing Beltline Weld Seam and Plate Locations. The Increase in the Axial Extent of the Beltline Requires that Radiation Damage Effects for Shell 1, 2 and 3 Materials be Included in the P-T Curve Analysis.

1-4

tvon-rroprtetary 2

NEUTRON TRANSPORT MODEL DESCRIPTION The neutron exposure of reactor structures is determined by a neutron transport calculation, or a combination of neutron transport calculations, to represent the distribution of neutron flux in three dimensions. The calculation determines the distribution of neutrons of all energies from their source from fission in the core region to their eventual absorption or leakage from the system. The calculation uses a model of the reactor geometry that includes the significant structures and geometrical details necessary to define the neutron environment at locations of interest. This chapter provides a summary of the neutron source representation used in the GGNS model along with a description of the key model features.

2.1 Source Representation During reactor operation, the neutron flux level at any point in the shroud or vessel will vary due to changes in fuel composition, power distributions within the core, and water void fraction. These changes occur between fuel cycles due to changes in fuel loading and fuel design, and within a fuel cycle due to fuel and poison bumup burnup and resultant changes in power shape, control rod position, fission contributions by nuclide, and void fraction vs. axial height in each fuel bundle. Power shape throughout a typical cycle's worth of operation has similar characteristics from cycle-to-cycle. Power starts out being preferentially produced in the bottom half of the core, and, as the fuel cycle progresses, the power peak shifts to higher core locations.

burnup decreases the fraction of fissions coming from U235 and increases the fraction The fuel bumup of fissions from plutonium isotopes. This results in slight changes in the fission spectrum and in the number of fissions per unit power. The control rod pattems patterns are altered at several discreet time intervals throughout the cycle. These sequence exchanges (SE) result in step changes in the power distribution and in the distribution of reactor water densities, especially in local areas near the control rods. [[

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Neutron Transport Model Description Non-]Jroprieta~

[[

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Neutron Transport Model Description Non-Proprietary

[[

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2.2 Neutron Transport Model The transport calculations for GGNS were carried out in 2Dand 3D geometries using the DaRT two-dimensional and the TORT three-dimensional discrete ordinates codes [2-13],

respectively. [[ ]] The DaRT code is an update of the DOT code which has been in use for this type of problem for many years. The TORT code has the same calculational methodology as DaRT, except that it extends the calculation into the third dimension. [[

]] The energy group boundaries for the 47 groups are given in Table 2-1. [[

]]

The computer codes were obtained from the Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory. Each code was then compiled on the computer used by MPM for the calculations and a series of test cases were run to verify the code performance. The test cases all agreed within allowable tolerance with established results. This verification was conducted under the MPM Nuclear Quality Assurance Program. The calculational procedures meet standards specified by the NRC and ASTM as appropriate. In particular, the analysis (including all modeling details and cross-sections) is consistent with RG 1.190 [2-16], and the 3D calculational methodology has been benchmarked to measured plant-specific BWR data as described in Section 3. [[

]]

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Neutron Transport Model Description Non-Proprietary

[[

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2D R-O Calculations The R-S layout at core axial midplane is shown schematically in Figure 2-1. Some of the structures in this figure (the surveillance capsule and jet pumps) are not to scale. A more detailed listing of dimensions for the various structures is given in Tables 2-2 through 2-4.

Dimensions were obtained from plant drawings and the GGNS design inputs which are referenced in the tablesthat were transmitted to MPM by Reference [2-19]. As shown in Figure 2-1, all structures outside the core were modeled with a cylindrical symmetry except for the inclusion of a surveillance capsule centered at 3 degrees and jet pump structures located in the downcomer region. [[

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Neutron Transport Model Description Non-Proprietary

[[

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2D R-Z Calculations

[[

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Flux Synthesis

[[ ]] In order to estimate the fluence rate in the three dimensional geometry, the following standard synthesis equation was used to evaluate the flux (qi) [[ ]]

<p(R,S,Z) = <peR,S) * <p(R,Z) / <peR) 2-5

Neutron Transport Model Description Non-Proprietary

[[

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3D TORT Calculations

[[

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Neutron Transport Model Description Non-Proprietary

[[

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2.3 Section 2 References

[2-1] [[

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.[2-2] RSICC Peripheral Shielding Routine Code Collection, PSR-277, LEPRICON, PWR Pressure Vessel Surveillance Dosimetry Analysis System, available from the Radiation Safety Information Computational Center, Oak Ridge National Laboratory, Oak Ridge, TN, June 1995.

[2-3] [[

]]

[2-4] [[

]]

[2-5] [[

]]

[2-6] [[

]]

[2-7] [[

]]

[2-8] [[

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[2-9] [[

]]

[2-10] [[

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[2-11] [[

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[2-12] [[

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Neutron Transport Model Description Non-Proprietary

[2-13] RSICC Computer Code Collection, CCC-543, TORT-DaRT-PC, Two- and Three-Dimensional Discrete Ordinates Transport Version 2.7.3, available from the Radiation Safety Information Computational Center, Oak Ridge National Laboratory, Oak Ridge, TN, June 1996.

[2-14] [[

]]

[2-15] ASTM Designation E482-89, Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1997.

[2-16] Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U. S. Nuclear Regulatory Commission, March 2001.

[2-17] [[

]]

[2-18] [[

]]

[2-19] [[

]]

2-8

Neutron Transport Model Description Non-Proprietary Table 2-1Neutron Energy Group Structure used in the GGNS Transport Calculations - 47 Groups.

Upper Energy Upper Energy Energy Group (MeV) Energy Group (MeV) 1 1.733E+01 25 2.972E-01 2 1.419E+01 26 1.832E-01 3 1.221E+01 27 1.111E-01 4 1.000E+01 28 6.738E-02 5 8.607E+00 29 4.087E-02 6 7.408E+00 30 3.183E-02 7 6.065E+00 31 2.606E-02 8 4.966E+00 32 2.418E-02 9 3.679E+00 33 2.188E-02 10 3.012E+00 34 1.503E-02 11 2.725E+00 35 7.102E-03 12 2.466E+00 36 3.355E-03 13 2.365E+00 37 1.585E-03 14 2.346E+00 38 4.540E-04 15 2.231E+00 39 2.145E-04 16 1.920E+00 40 1.013E-04 17 1.653E+00 41 3.727E-05 18 1.353E+00 42 1.068E-05 19 1.003E+00 43 5.044E-06 20 8.208E-01 44 1.855E-06 21 7.427E-01 45 8.764E-07 22 6.081E-Ol 46 4.140E-07 23 4.979E-Ol 47 1.000E-07 24 3.688E-01 2-9

Neutron Transport Model Description Non-]Jroprieta~

Table 2-2 GGNS Reactor Component Radial Dimensions.

[[

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2-10

Neutron Transport Model Description Non-Proprietary Table 2-2 GGNS Reactor Component Radial Dimensions (continued).

[[

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2-11

Neutron Transport Model Description Non-Proprietary Table 2-2GGNS Reactor Component Radial Dimensions (continued).

[[

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Neutron Transport Model Description Non-Proprietary Table 2-3GGNS Reactor Component Azimuthal Locations.

[[

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2-13

Neutron Transport Model Description Non-Proprietary Table 2-4GGNS Reactor Component Axial Dimensions.

[[

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2-14

Neutron Transport Model Description Non-Proprietary

[[

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Figure 2-1GGNS R-O Geometry used in the TORT Calculations.

2-15

Neutron Transport Model Description Non-Proprietary 150 TAJ?

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2-16

tvon-rroprtetary BENCHMARK ANALYSES NRC Regulatory Guide 1.190 requires that neutron transport methods satisfy compliance requirements and they must also demonstrate specified accuracy through benchmarking. This section of the report summarizes the work done to demonstrate that the MPM methods meet the NRC's requirements.

3.1 Compliance with RG 1.190 The United States Nuclear Regulatory Commission has issued RG 1.190 entitled, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" [3-1].

This guide covers recommended practices for neutron transport calculations and applies to other reactor components in addition to the primary emphasis on the pressure vessel. The regulatory positions in the guide that pertain to calculational methodology are summarized in Table 3-1 which is taken directly from RG 1.190. The table references paragraphs in the guide that give more detailed information on each position. The compliance of the GGNS fluence calculations with the guide is summarized below.

Fluence Calculational Methods Fluence Determination: This calculation was performed using an absolute fluence calculation.

Meets guide requirement.

Modeling Data: The MPM methodology is based on documented and verified plant-specific data. Further, the calculations use as-built data for plant structures and material compositions whenever these data are available. The fuel data is specific for each fuel cycle and includes results for power distributions and water densities taken from the fuel depletion analysis.

Meets guide requirement.

Nuclear Data: The calculations used the [[ ]]cross section set that is based on the latest version (VI) of the Evaluated Nuclear Data File (ENDF/B). The [[ ]] set has undergone extensive testing and benchmarking to ensure its validity for light water reactor (LWR) calculations.

Meets guide requirement.

Cross-Section Angular Representation: The calculations use a P3 angular decomposition in accordance with the guide.

Meets guide requirement.

Cross-Section Group Collapsing: All calculations are performed with the [[ ]] library which is collapsed to 47 neutron groups. Benchmarking has shown that the 47-group structure is adequate for L WR neutron transport calculations.

Meets guide requirement.

Neutron Source: The neutron source is calculated taking into account changes in neutrons per fission, energy per fission, and the average fission spectrum which develops as the U235 is 3-1

Benchmark Analyses JVon-]Jroprieta~

burned and other isotopes, such as Pu239, increase in fission fraction.

Meets guide requirement.

End-of-Life Predictions: The fluence has been calculated for every cycle up to the present. [[

]]

Meets guide requirement.

Spatial Representation: The present methodology meets or exceeds these requirements. [[

]]

Meets guide requirement.

Multiple Transport Calculations: It was not necessary to use bootstrapping for these calculations so this requirement does not apply.

Point Estimates: This requirement only applies to Monte Carlo calculations which are not used here.

Statistical Tests: This requirement only applies to Monte Carlo calculations which are not used here.

Variance Reduction: This requirement only applies to Monte Carlo calculations which are not used here.

Capsule Modeling: The capsule geometry is modeled using as-built drawings. [[

]].

Meets guide requirement.

3-2

Benchmark Analyses Non-Proprietary Spectral Effects on RTNDT: This requirement only applies to extrapolation through the vessel and does not affect the benchmark calculations. However, when fluence within the vessel is required, the displacement per atom (dpa) extrapolation methodology is applied to vessel calculations as specified in RG 1.99, Revision 2. Data are supplied to enable extrapolation using dpa calculated extrapolation or using RG 1.99 Rev. 2 extrapolation to account for the spectral shift.

Meets guide requirement.

Cavity Calculations: Cavity calculations are not important at GGNS, and no cavity dosimetry work has been performed at GGNS.

Methods Qualification: Methods qualification for these calculations is discussed in detail in Sections 3.2 and 3.3, which deals with benchmarking of the methodology. A complete analytical uncertainty analysis is described in Section 6, and this analysis was carried out in accordance with the guide. In summary, an extensive benchmarking program has been carried out to qualify the MPM neutron transport methodology. All of the requirements of RG 1.190 have been met.

In particular, all CIM results fall within allowable limits (+/--20%), and it was determined that no bias need be applied to MPM fluence results. The uncertainty analysis indicates that all fluence results in the beltline region have uncertainty of less than 20%.

Meets guide requirement.

Fluence Calculational Uncertainty: An extensive evaluation of all contributors to the uncertainty in the calculated fluence was made for the BWR plant calculations performed to date. This evaluation indicated that the uncertainty in calculated fluences in the reactor beltline region is below 20% as specified in the guide. In addition, the comparisons with measurements indicate agreement well within the 20% limit. The agreement of calculations with measurements to within +/-20% uncertainty indicates that the MPM calculations can be applied for fluence determination with no bias.

Meets guide requirement.

Fluence Measurement Methods The regulatory positions in the guide that pertain to fluence measurement methods are summarized in Table 3-2.The compliance of the GGNS fluence measurements with the guide is summarized below.

Spectrum Coverage: GGNS does not have dosimetry sets installed to provide spectrum definition. This is in common with GE BWRs which have only limited dosimetry installed in surveillance capsules. The GGNS dosimetry analyzed to date is from iron wires attached to a surveillance capsule and removed after the first cycle of operation. Calculated neutron spectra are validated using the detailed measurements for test reactors and the calculational benchmark included in RG 1.190. Results are documented in Reference [3-2].

Dosimeter Nuclear and Material Properties: The GGNS dosimetry wire material characterization was performed by GE (Reference [3-3 J). All nuclear constants and parameters used in the dosimetry counting and analysis follow ASTM standard procedures and use validated nuclear constants.

Meets guide requirement.

3-3

Benchmark Analyses Non-Proprietary Corrections: [[

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Meets guide requirement.

Response Uncertainty: Uncertainty analyses for each dosimeter have been reported in all of the surveillance capsule reports performed by MPM. [[

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Meets guide requirement.

Validation: [[

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Fast-Neutron Fluence: M/C ratios are determined for each dosimeter measurement using calculated detector responses (Reference [3-2]).

Meets guide requirement.

Measurement-to-Calculated Ratios: The M/C ratios, standard deviations, and comparisons between calculation and measurement have been done and have been reported as discussed earlier (References [3-2]).

Meets guide requirement.

Reporting Provisions The regulatory positions in the guide that pertain to neutronfluence and uncertainty reporting are summarized in Table 3-3.The compliance of the GGNS reporting with the guide is summarized below.

Neutron Fluence and Uncertainties: This report on the GGNS calculations, as well as the benchmark report (Reference [3-2]), meets all RG 1.190 reporting requirements.

Meets guide requirement.

Multigroup Fluences: Because of the extensive amount of data, only the fluence distributions for neutrons with energy greater than 1 MeV are reported, except for the vessel where neutrons 3-4

Benchmark Analyses Non-Proprietary with energy greater than 0.1 MeV and dpa are also reported. The multi group fluence rates are permanently stored for future access if required. These are not seen as having any application except for possible future dosimetry analysis. Multigroup flux values are readily available for all dosimetry locations.

Meets guide requirement.

Bias Reporting: No bias is observed and none is applied.

Meets guide requirement. .

Integral Fluences and Uncertainties: These are all reported.

Meets guide requirement.

Comparisons ofCalculation and Measurement: These are reported for all dosimetry locations.

Meets guide requirement.

Standard Neutron Field Validation: Not applicable.

Specific Activities and Average Reaction Rates: This is contained in all reports with measured dosimetry data.

Meets guide requirement.

Corrections and Adjustments to Measured Quantities: No corrections made.

Meets guide requirement.

3.2 Summary of MPM Methods The neutron exposure of reactor structures to determine the spatial distribution of neutrons in the reactor components is determined by either a 3D neutron transport calculation, or by a combination ofR-a, R-Z, and R neutron transport calculations in support of the 2D synthesis method. In particular, the calculation determines the distribution of neutrons of all energies from their source from fission in the core region to their eventual absorption or leakage from the system. The transport calculations use a model of the reactor geometry that includes the significant structures and geometrical details necessary to accurately define the neutron environment at all locations of interest. A brief summary of the MPM methods is included here.

[[

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Whenever possible, the plant geometry is modeled using as-built drawings. In cases where as-built drawings are not available, design drawings are used. During reactor operation, the neutron flux level at any point in the shroud or vessel will vary due to changes in fuel composition, power distributions within the core, and water void fraction. These changes occur between fuel cycles due to changes in fuel loading and fuel design, and within a fuel cycle due to fuel and poison bumup and resultant changes in power shape, control rod position, fission contributions by nuclide, and void fraction vs. axial height in each fuel bundle. To achieve the 3-5

Benchmark Analyses lVon-JDroprieta~

highest accuracy, changes in the three-dimensional bundle power distributions and water density distributions must be appropriately modeled. [[

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3.3 Benchmarking MPM Methodology Although RG 1.190 was specifically developed to address calculation of fluence to the vessel, the guide can be considered to apply to other reactor components such as the shroud, surveillance capsules, and internal structures. As mentioned, MPM methods include the standard 2D synthesis, [[ ]] and 3D TORT.In order to meet the methods qualification requirement of RG 1.190, the MPM calculational methodology has been validated by comparison with measurement and calculational benchmarks. Comparisons of calculations with measurements have been made for the Pool Critical Assembly (PCA) pressure vessel simulator benchmark, [[

]] Comparisons with a BWR calculational benchmark have also been completed.

The qualification of the methods used for the reactor transport calculations can be divided into two parts. The first part is the qualification of the cross-section library and calculational methods that are used to calculate the neutron transport. The second part is the validation of the method by comparison of calculations with dosimetry measurements for both representative plants and also for the current plant being analyzed, which, in the current case, is the GGNS plant. Details of the extensive benchmarking efforts performed by MPM are contained in a separate document [3-2].

Measurement and Calculational Benchmarks The qualification of the cross-section library and calculational methods is particularly important for vessel fluence calculations because of the large amount of neutron attenuation between the source in the core and the vessel. The cross-sections are first developed as an evaluated file that details all of the reactions as a continuous function of energy. The ENDF/B evaluators take into account various measurements, including integral measurements (such as criticality and dosimetry measurements) that provide a test of the adequacy of the evaluation.

F9r transport calculations using discrete ordinates, the ENDF/B cross-section files must be collapsed into multi group files. This is done in two steps. First, fine-group cross-sections are calculated (Vitamin B6 library). [[ ]]

3-6

Benchmark Analyses Non-Proprietary contains cross-sections collapsed using a BWR core spectrum, a PWR core spectrum, a PWR downcomer spectrum, a PWR vessel spectrum, and a concrete shield spectrum. These various cross-section libraries are then tested against various benchmarks and compared with measured results and results calculated using the fine-group cross-sections [3-7]. [[

]]

[[

[[

]] A particularly appropriate benchmark for geometry outside the core which includes the vessel, is the PCA benchmark [3-9]. This benchmark provides validation of the transport through typical reactor structures and a simulated reactor vessel in a simple geometry.

The PCA has high-accuracy measurement results extending from inside a simulated thermal shield through to the outside of a simulated vessel. The PCA benchmark was calculated using the MPM methodology and detailed results are reported in [3-2]. [[

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Another benchmarking requirement of RG 1.190 is to compare with a suitable calculational benchmark. The calculational benchmarks used to satisfy this requirement are documented in Reference [3-11]. The benchmark problems include 3 different PWR geometries and a single BWR problem. It is intended that the analyst select the benchmark problem or problems appropriate to the plant being analyzed. [[

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Benchmark Analyses Non-Proprietary Benchmarking MPM Calculations against Measured Data in Operating BWRs The second part of benchmarking is to compare calculations with measurements in a geometry as close as possible to that which is being analyzed. Comparisons were made with BWR surveillance dosimetry measurements [[

]]

Calculated activation results from each dosimeter type all fell within +/-20% of the measurement. The dosimetry results have been averaged for each set and the results are shown in Table 3-6. If the average results from the nine dosimetry sets are averaged, the average calculated-to-measured (C/M) ratio [[ . ]] which indicates that the MPM methodology does not exhibit any consistent bias for BWR calculations. This meets the criterion set by RG 1.190 for acceptability of the calculations.

Benchmarking 3D TORT Methods

[[

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Benchmark Analyses Non-Proprietary

[[

]] very good agreement between the 3D calculation and the measurement,

[[

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3.4 Conclusions In summary, it is concluded that the RG 1.190 requirement for qualification of the MPM methods for BWR neutron transport analyses by comparisons to measurement and calculational benchmarks has been fully satisfied. Moreover, the agreement of calculations with measurements to within +/-20% uncertainty indicates that the MPM calculations can be applied for fluence determination with no bias.

3.5 Section 3 References

[3-1] Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U. S. Nuclear Regulatory Commission, March 2001.

[3-2] [[

]]

[3-3] [[

]]

[3-4] W. N. McElroy, "Data Development and Testing for Fast Reactor Dosimetry ", Nucl.

Tech. 25, 177 (1975).

[3-5] R. Gold and W. N. McElroy, "The Light Water Reactor Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP): Past Accomplishments, Recent Developments, and Future Directions," Reactor Dosimetry: Methods, Applications, and Standardization, ASTM STP 1001, Harry Farrar and E. P. Lippincott, eds., American Society for Testing and Materials, Philadelphia, 1989, pp 44-61.

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[3-6] [[

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[3-7] [[ ]]

[3-8] [[

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[3-9] . Remec, 1. and Kam, F.B.K., Pool Critical Assembly Pressure Vessel Facility Benchmark, NUREG/CR-6454, (ORNL/TM-13205), USNRC, July 1997.

NUREGICR-6454,

[3-10] [[

]]

[3-11] Carew, J. F., Hu, K., Aronson, A., Prince, A., and Zamonsky, G., PWR and BWR Pressure Vessel Fluence Calculational Benchmark Problems and Solutions, NUREGICR-NUREG/CR-6115 (BNL-NUREG-52395), September, 2001.

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Benchmark Analyses Non-Proprietary Table 3-1 Summary of Regulatory Guide 1.190 Positions on Fluence Calculation Methods.

Regulatory Regulatory Position - Fluence Calculation Methods Section Fluence Determination. Absolute fluence calculations, rather than extrapolated fluence 1.3 measurements, must be used for the fluence determination.

Modeling Data. The calculation modeling (geometry, materials, etc.) should be based on 1.1.1 documented and verified plant-specific data.

Nuclear Data. The latest version of the Evaluated Nuclear Data File (ENDF/B) should 1.1.2 be used for determining nuclear cross- sections. Cross-section sets based on earlier or equivalent nuclear-data sets that have been thoroughly benchmarked are also acceptable.

When the recommended cross-section data change, the effect of these changes on the licensee-specific methodology must be evaluated and the fluence estimates updated when the effects are significant.

Cross-Section Angular Representation. In discrete ordinates transport calculations, a P3 1.1.2 angular decomposition of the scattering cross-sections (at a minimum) must be employed.

Cross-Section Group Collapsing. The adequacy of the collapsed job library must be 1.1.2 demonstrated by comparing calculations for a representative configuration performed with both the master library and the job library.

Neutron Source. The core neutron source should account for local fuel isotopics and, 1.2 where appropriate, moderator density. The neutron source normalization and energy dependence must account for the fuel exposure dependence of the fission spectra, the number of neutrons produced per fission, and the energy released per fission.

End-of-Life Predictions. Predictions of the vessel end-of-life fluence should be made 1.2 with a best-estimate or conservative generic power distribution. If a best estimate is used, the power distribution must be updated if changes in core loadings, surveillance measurements, or other information indicate a significant change in projected fluence values.

Spatial Representation. Discrete ordinates neutron transport calculations should 1.3.1 incorporate a detailed radial- and azimuthal-spatial mesh of -2 intervals per inch radially. The discrete ordinates calculations must employ (at a minimum) an S8 quadrature and (at least) 40-80 intervals per octant.

Multiple Transport Calculations. If the calculation is performed using two or more 1.3.1 "bootstrap" calculations, the adequacy of the overlap regions must be demonstrated.

3-11

Benchmark Analyses Non-Proprietary Table 3-1 Summary of Regulatory Guide 1.190 Positions on Fluence Calculation Methods (continued).

Regulatory Regulatory Position - Fluence Calculation Methods Section Point Estimates. If the dimensions of the tally region or the definition of the average- 1.3.2 flux region introduce a bias in the talley edit, the Monte Carlo prediction should be adjusted to eliminate the calculational bias. The average-flux region surrounding the point location should not include material boundaries or be located near reflecting, periodic or white boundaries.

Statistical Tests. The Monte Carlo estimated mean and relative error should be tested 1.3.2 and satisfy all statistical criteria.

Variance Reduction. All variance reduction methods should be qualified by comparison 1.3.2 with calculations performed without variance reduction.

Capsule Modeling. The capsule fluence is extremely sensitive to the geometrical 1.3.3 representation of the capsule geometry and internal water region, and the adequacy of the capsule representation and mesh must be demonstrated.

Spectral Effects on RT NOT. In order to account for the neutron spectrum dependence of 1.3.3 RTNOT, when it is extrapolated from the inside surface of the pressure vessel to the T/4 and 3T/4 vessel locations using the> I-MeV fluence, a spectral lead factor must be .

applied to the fluence for the calculation of ~RTNOT.

Cavity Calculations. In discrete ordinates transport-calculations, the adequacy of the S8 1.3.5 angular quadrature used in cavity transport calculations must be demonstrated.

Methods Qualification. The calculational methodology must be qualified by both (1) 104.1,104.2, comparisons to measurement and calculational benchmarks and (2) an analytic 104.3 uncertainty analysis. The methods used to calculate the benchmarks must be consistent (to the extent possible) with the methods used to calculate the vessel fluence. The overall calculational bias and uncertainty must be determined by an appropriate combination of the analytic uncertainty analysis and the uncertainty analysis based on the comparisons to the benchmarks.

Fluence Calculational Uncertainty. The vessel fluence (1 sigma) calculational 1,1.4.3 uncertainty must be demonstrated to be 200/0 for RTpTs and RTNOT determination. In these applications, if the benchmark comparisons indicate differences greater than -20%,

the calculational model must be adjusted or a correction must be applied to reduce the difference between the fluence prediction and the upper l-sigma limit to within 20%.

For other applications, the accuracy should be determined using the approach described in Regulatory Position lA, and an uncertainty allowance should be included in the fluence estimate as appropriate in the specific application.

3-12

Benchmark Analyses Non-Proprietary Table 3-2 Summary of Regulatory Guide 1.190 Positions on Fluence Measurement Methods.

Regulatory Regulatory Position - Fluence Measurement Methods Section Spectrum Coverage. The set of dosimeters should provide adequate spectrum coverage. 2.1.1 Dosimeter Nuclear and Material Properties. Use of dosimeter materials should address 2.1.1 melting, oxidation, material purity, total and isotopic mass assay, perturbations by encapsulations and thermal shields, and accurate dosimeter positioning. Dosimeter half-life and photon yield and interference should also be evaluated.

Corrections. Dosimeter-response measurements should account for fluence rate 2.1.2 variations, isotopic burnup effects, detector perturbations, self-shielding, reaction interferences, and photofission.

Response Uncertainty. An uncertainty analysis must be performed for the response of 2.1.3 each dosimeter.

Validation. Detector-response calibrations must be carried out periodically in a standard 2.2 neutron field.

Fast-Neutron Fluence. The E > 1 MeV fast-neutron fluence for each measurement 2.3 location must be determined using calculated spectrum-averaged cross-sections and individual detector measurements. As an alternative, the detector responses may be used to determine reaction probabilities or average reaction rates.

Measurement-to-Calculation Ratios. The M/C ratios, the standard deviation and bias 2.3 between calculation and measurement, must be determined.

3-13

Benchmark Analyses Non-Proprietary Table 3-3 Summary of Regulatory Guide 1.190 Positions on Fluence and Uncertainty Reporting.

Regulatory Regulatory Position - Reporting Provisions Section Neutron Fluence and Uncertainties.Details of the absolute fluence calculations, 3.1 associated methods qualification and fluence adjustments (if any) should be reported.

Justification and a description of any deviations from the provisions of this guide should be provided.

Multigroup Fluences.Calculated multigroup neutron fluences and fluence rates should be 3.2 reported.

Bias Reporting.The value and basis of any bias or model adjustment made to improve 3.2 the measurement-to-calculation agreement must be reported.

Integral Fluences and Uncertainties.Calculated integral fluences and fluence rates for E > 3.3 1 MeV and their uncertainties should be reported.

Comparisons of Calculation and Measurement.Measured and calculated integral E > 1 3.4 MeV fluences or reaction rates and uncertainties for each measurement location should be reported. The M/C ratios and spectrum averaged cross-section should also be reported.

Standard Neutron Field Validation.The results of the standard field validation of the 3.5 measurement method should be reported.

Specific Activities and Average Reaction Rates.The specific activities at the end of 3.5 irradiation and measured average reaction rates with uncertainties should be reported.

Corrections and Adjustments to Measured Quantities.All corrections and adjustments to 3.5 the measured quantities and their justification should be reported.

3-14

Benchmark Analyses Non-Proprietary Table 3-4 Comparison of Calculated and Measured Results for peA.

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3-15

Benchmark Analyses Non-Proprietary Table 3-5 Comparison of Calculated and Benchmark Results for the Reactor Vessel Calculated at Reactor Axial Peak.

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3-16

Benchmark Analyses Non-Proprietary Table 3-6 Tabulation of Dosimetry Results for Operating BWR Plants.

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3-17

ivon-rroprtetary SHROUD AND TOP GUIDE FLUENCE RESULTS l An important concern for many BWR reactors is the neutron exposure of the shroud and top guide welds in these components. Because the shroud is located close to the fuel region, the exposure is relatively high, and this must be taken into account to evaluate the possibility of stress corrosion cracks and crack growth. Of special importance is determining weld locations that exceed the irradiation assisted stress corrosion cracking (lASCC) fast (E> 1.0 MeV) neutron threshold fluence. A fluence of 5E+20 n/crrr' has been proposed as a screening threshold, but more recent data suggest a lower threshold for some materials. In any event, radiation damage effects on the SCC crack growth model must be considered in setting future inspection intervals.

4.1 Shroud Weld Fluence Results Locations of all the shroud vertical and horizontal welds are shown schematically in Figures 1-1 and 1-2, and the weld locations are listed in Tables4-1 and 4-2. Evaluations of the fluence for all of these welds have been performed using the 3D TORT method described in Section 2. Tables 4-3 and 4-4 summarize the calculated maximum fluences to the shroud and top guide welds. These fluences were calculated at the inner diameter (ID) surface of the welds at the point along the welds where the fluence is a maximum. [[

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4-1

Shroud and Top Guide Fluence Results Non-Proprietary

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4-2

Shroud and Top Guide Fluence Results Non-Proprietary

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4.2 Comparison with Past Results

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4-3

Shroud and Top Guide Fluence Results Non-Proprietary

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4.3 Section4 References

[4-1] [[

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[4-2] [[

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4-4

Shroud and Top Guide Fluence Results Non-Proprietary Table 4-1 Locations for Fluence Evaluation in the Shroud and Top Guide Horizontal Welds.

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Table 4-2 Locations for Fluence Evaluation in the Shroud and Top Guide Vertical Welds.

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4-5

Shroud and Top Guide Fluence Results Non-Proprietary Table 4-3 Estimated Maximum Fluence to GGNS Shroud and Top Guide Horizontal Welds.

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4-6

Shroud and Top Guide Fluence Results JVon-j>roprieta~

Table 4-4 Estimated Maximum Fluence to GGNS Shroud and Top Guide Vertical Welds.

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4-7

Shroud and Top Guide Fluence Results Non-Proprietary

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Figure 4-1Fast Flux (E>1.0 MeV) at the Shroud IR to Weld H4 for Cycles 1 through 21.

4-8

Shroud and Top Guide Fluence Results Non-Proprietary

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Figure 4-2 Fast Fluence (E>1.0 MeV) to Shroud Weld H4 at the End of Cycle 21 (28.088 EFPY).

4-9

Shroud and Top Guide Fluence Results Non-Proprietary

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Figure 4-3 Fast Fluence (E>l.O MeV) to Shroud Weld H4 at 35 EFPY.

4-10

Shroud and Top Guide Fluence Results Non-Proprietary

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Figure 4-4 Fast Fluence (E>1.0 MeV) to Shroud Weld H4 at 54 EFPY.

4-11

Shroud and Top Guide Fluence Results Non-Proprietary

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Figure 4-5 Fast Fluence (E>l.O MeV) to Shroud Vertical Welds V13-V16 at the End of Cycle 21 (28.088 EFPY).

4-12

Shroud and Top Guide Fluence Results Non-Proprietary

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Figure 4-6 Fast Fluence (E>1.0 MeV) to Shroud Vertical Welds V13-V16 at 35 EFPY.

4-13

Shroud and Top Guide Fluence Results Non-Proprietary

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Figure 4-7 Fast Fluence (E>l.O MeV) to Shroud Vertical Welds V13-V16 at 54 EFPY.

4-14

tvon-rroprtetary 5

VESSEL AND CAPSULE FlUENCE RESULTS 5.1 Cycle 1 Dosimetry Analysis The GGNS cycle 1 dosimetry was originally analyzed using 2Dsynthesis [5-1]. The past work has been reanalyzed using 3D TORT calculations. [[

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The detailed power history for GGNS cycle 1 operation is presented in Table 5-3 [5-2].

[[

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The removable dosimetry packet consisted of 3 iron wires. The dosimetry packet was located on the side of the capsule. The location of the dosimetry was provided in the plant documentation as indicated in Table 2-3. [[

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5-1

Vessel and Capsule Fluence Results Non-Proprietary

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5.2 Pressure Vessel Analysis The analysis of the vessel was carried out using the 3D methodology described in Section

2. The maximum flux location at the vessel shell course plates varies slightly from cycle-to-cycle as a result of the different fuel loadings. Therefore, the maximum fluence for the shell course plates was conservatively calculated [[

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Plots of the azimuthal and axial traverses of fluence at the vessel wetted surface at the end of cycle 21, along with the results at the vessel 1/4T and 3/4T positions, are shown in Figures 5-1 and 5-2, respectively. [[

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The NRC defines the beltline region in 10CFR50, Appendix G as "the region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage." [[

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5-2

Vessel and Capsule Fluence Results Non-Proprietary

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As shown in Figure 5-3, the reactor vessel is made up of plates that are welded together.

For the purpose of PT curve calculations, fluences have been evaluated for shell I, shell 2, and shell 3 materials. [[

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Radiation embrittlement effects are often correlated with fast fluence (E > 1 MeV).

However, it is generally thought that dpa might be a better correlation parameter since it accounts for spectral effects and, if this is correct, the use of the fast fluence (E > 1 MeV) values within the vessel might under predict the radiation damage at locations within the vessel.

Therefore, the NRC requires the use of a dpa attenuation in the vessel for preparation of 5-3

Vessel and Capsule Fluence Results Non-Proprietary pressure-temperature operating curves. The fluence attenuation factors within the vessel can be evaluated using calculated dpa attenuation from Table 5-9 or using the dpa formulation specified in the RG 1.99 (Rev 2) [5-6]. The fluence values using both these attenuation methods are given in Table 5-18[[

]]This evaluation of the dpa cross section is based on the ENDF-IV cross section file.[[

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The vessel has nozzle penetrations at several locations, and neutron exposure at the nozzles is of concern for neutron damage analysis. Four sets of nozzles were evaluated, and the locations of these nozzles along with the fluence evaluation points are summarized in Table 5-

19. As shown in the table, the nozzle fluences are determined at the maximum point in the nozzle weld as well as at the conservative locations specified in Reference [5-9]. For nozzles below the core, the maximum fluence point occurs at the top of the nozzle. The reverse is true for nozzles above the core. The maximum fluence values at the nozzles are summarized in Table 5-20. Fluences at conservative locations near the nozzles specified by GGNS in Reference [5-9]

are given in Tables 5-21 and 5-22. The data in Table 5-21 is at the elevation specified, but at the maximum azimuthally, whereas the data in Table 5-22 is at the locations specified in [5-9]. [[

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Section 6 presents the N6 and N12 maximum nozzle fluence uncertainties. The N12 nozzle elevation is close to that of the active fuel, [[

]] Regulatory Guide 1.190 states that if the analytical uncertainty at the 1 sigma level is greater than 30%, the methodology of the regulatory guide is not applicable and the plant application will be reviewed on an individual basis. [[

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5-4

Vessel and Capsule Fluence Results Non-Proprietary

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5.3 Comparison of RPV Peak Fluence with Past Results

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5.4 RPV Internals Analysis In addition to the fluence analysis for the RPV discussed previously, calculation of fluences at the nineteen locations specified by GEH in Table 5-24 was completed in support of the RPV internals mechanical evaluation. Table 5-24 specifies radial positions and axial elevations, but no angular data were specified. This is because GEH uses the peak values for the structures and components specified to ensure conservatism in the calculations. In the case of the shroud, the peak on the ID surface of the shroud is reported. For plates, such as the core plate, the peak anywhere on the plane defining the plate has been calculated. For components with discrete angular positions, such as the shroud support legs, the peak reported is the peak over the 360 degrees at the elevation of the legs. Thus, in these cases, the peak will not necessarily be located at the component. For the components with specified axial ranges, such as the Core Plate Bolts and the Top Guide Bolts, GEH has requested the peak as just defined over 360 degrees, as well as the average over the axial extent given in Table 5-24, at the peak azimuthal location.

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5-5

Vessel and Capsule Fluence Results Non-Proprietary 5.5 Section 5 References

[5-1] [[

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[5-2] [[

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[5-3] ASTM Designation E263-00, Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 2003.

[5-4] ASTM Designation E1005-03, Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E706(IIIA), in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 2003.

[5-5] [[

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[5-6] Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U. S. Nuclear Regulatory Commission, May 1988.

[5-7] ASTM Designation E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E706(ID), in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,2000.

[5-8] ASTM Designation E693-01, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E706(ID), in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,2003.

[5-9] [[

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5-6

Vessel and Capsule Fluence Results Non-Proprietary Table 5-1Calculated Fe(n,p) Reaction Rates at the Cycle 1 Dosimetry Location.

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5-7

Vessel and Capsule Fluence Results Non-Proprietary Table 5-2 Flux Spectrum at Dosimetry Location.

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5-8

Vessel and Capsule Fluence Results Non-Proprietary Table 5-3 Grand Gulf Cycle 1 Power History.

Period Period Average Cycle Cumulative Effective FuJI Power Fraction Of Full Date Days In Period (MWdlMTU) Days Power 30-Sep-84 Begin 0 3-Nov-84 35 178 6.789 0.1940 14-Nov-84 11 235 2.174 0.1976 21-Nov-84 7 295 2.288 0.3269 4-Feb-85 75 618 12.320 0.1643 12-Apr-85 67 906 10.985 0.1640 25-Apr-85 13 1071 6.293 0.4841 17-May-85 22 1384 11.938 0.5427 4-Jun-85 18 1666 10.756 0.5976 3-Jul-85 29 2140 18.079 0.6234 15-Jul-85 12 2351 8.048 0.6707 26-Jul-85 11 2596 9.345 0.8495 7-Aug-85 12 2863 10.184 0.8487 21-Aug-85 14 3158 11.252 0.8037 23-Aug-85 2 3187 1.106 0.5531 28-Aug-85 5 3269 3.128 0.6255 5-Sep-85 8 3466 7.514 0.9392 10-Sep-85 5 3591 4.768 0.9535 19-5ep-85 9 3808 8.277 0.9196 22-Sep-85 3 3865 2.174 0.7247 26-Sep-85 4 3914 1.869 0.4672 30-Sep-85 4 4026 4.272 1.0680 12-0ct-85 12 4334 11.748 0.9790 15-Dec-85 64 4360 0.992 0.0155 18-Dec-85 3 4459 3.776 1.2587 28-Dec-85 10 4586 4.844 0.4844 9-Jan-86 12 4861 10.489 0.8741 ll-Jan-86 2 4891 1.144 0.5721 31-Jan-86 20 5177 10.909 0.5454 12-Feb-86 12 5376 7.590 0.6325 25-Feb-86 13 5498 4.653 0.3579 28-Feb-86 3 5551 2.022 0.6738 15-Mar-86 15 5812 9.955 0.6637 18-Apr-86 34 6250 16.706 0.4914 28-Apr-86 10 6419 6.446 0.6446 5-9

Vessel and Capsule Fluence Results Non-Proprietary Table 5-3 Grand Gulf Cycle 1 Power History (continued).

Period Period Average Cycle Cumulative Effective Full Fraction of Full Date Days in Period (MWd/MTU) Power Days Power 9-May-86 11 6612 7.361 0.6692 25-May-86 16 6907 11.252 0.7032 28-May-86 3 6964 2.174 0.7247 13-Jun-86 16 7286 12.282 0.7676 27-Jun-86 14 7615 12.549 0.8963 3-Jul-86 6 7738 4.691 0.7819 18-Jul-86 15 8038 11.442 0.7628 30-Jul-86 12 8229 7.285 0.6071 7-Aug-86 8 8389 6.103 0.7628 18-Aug-86 11 8586 7.514 0.6831 23-Aug-86 5 8701 4.386 0.8773 3-Sep-86 11 8782 3.089 0.2809 5-Sep-86 2 8823 1.564 0.7819 Table 5-4Nuclear Parameters Used in the Evaluation of Neutron Sensors.

Approximate Monitor Isotopic Response Product Material Reaction of Interest Fraction Threshold Half-Life Iron Fe 54(n,p)Mn54 0.05845 2 MeV 312.3 days Table 5-5Tabulation of Dosimetry Results.

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5-10

Vessel and Capsule Fluence Results Non-Proprietary Table 5-6 Azimuthal.Variation of Maximum Fluence at the Vessel Wetted Surface.

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5-11

Vessel and Capsule Fluence Results Non-Proprietary Table 5-6Azimuthal Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-12

Vessel and Capsule Fluence Results Non-Proprietary Table 5-6 Azimuthal Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-13

Vessel and Capsule Fluence Results Non-Proprietary Table 5-6 Azimuthal Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-14

Vessel and Capsule Fluence Results Non-Proprietary Table 5-6 Azimuthal Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-15

Vessel and Capsule Fluence Results Non-Proprietary Table 5-6 Azimuthal Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-16

Vessel and Capsule Fluence Results Non-Proprietary Table 5-6 Azimuthal Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-17

Vessel and Capsule Fluence Results Non-Proprietary Table 5-7 Axial Variation of Maximum Fluence at the Vessel Wetted Surface.

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5-18

Vessel and Capsule Fluence Results Non-Proprietary Table 5-7 Axial Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-19

Vessel and Capsule Fluence Results Non-Proprietary Table 5-7 Axial Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-20

Vessel and Capsule Fluence Results Non-Proprietary Table 5-7 Axial Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-21

Vessel and Capsule Fluence Results Non-Proprietary Table 5-7 Axial Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-22

Vessel and Capsule Fluence Results Non-Proprietary Table 5-7 Axial Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-23

Vessel and Capsule Fluence Results

!jon-Proprietary Table 5-7 Axial Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-24

Vessel and Capsule Fluence Results Non-Proprietary Table 5-7 Axial Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-25

Vessel and Capsule Fluence Results Non-Proprietary Table 5-7Axial Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

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5-26

Vessel and Capsule Fluence Results Non-Proprietary Table 5-8 Relative Exposure through Vessel at Maximum Fluence Point at the End of Cycle 21 (28.088 EFPY).

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5-27

Vessel and Capsule Fluence Results Non-Proprietary Table 5-9 Calculated Maximum Vessel Exposure (Shell 2) at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures.

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5-28

Vessel and Capsule Fluence Results Non-Proprietary Table 5-10 Azimuthal Location Ranges of Plates in Vessel Shell 1, Shellz, and Shell

3. I Azimuthal Equivalent Azimuthal Angle Angle Range in First Plate ID Range Quadrant (degrees) (degrees)

Shell 1 PC MK 21-1-1 0-120 0-90 Plate Heat All13 PC MK 21-1-2 120-240 0-60 Plate Heat C2557 PC MK 21-1-3 240-360 0-90 Plate Heat C2506 Shell 2 PC MK 22-1-1 322-52 0-52 Plate Heat C2593 PC MK22-1-2 52-142 38-90 Plate Heat C2594 PC MK 22-1-3 142-232 0-52 Plate Heat C2594 PC MK 22-1-4 232-322 38-90 Plate HeatA 1224 Shell 3 PC MK 23-1-1 349-109 0-90 Plate Heat C2741 PC MK 23-1-2 109-229 0-71 Plate Heat C2779 PC MK 23-1-3 229-349 11-90 Plate Heat C2741 5-29

Vessel and Capsule Fluence Results Non-Proprietary Table 5-11 Calculated Maximum Vessel Shell 1 Exposure at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures.

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5-30

Vessel and Capsule Fluence Results JVon-]Jroprieta~

Table 5-12 Calculated Maximum Vessel Shell 3 Exposure at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures.

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5-31

Vessel and Capsule Fluence Results Non-Proprietary Table 5-13 Azimuthal Locations of Vertical Welds in Vessel Shell 1, Shell 2 and Shell 3.

Azimuthal Equivalent Azimuthal Weld ID Angle Angle in First Quadrant (degrees) (degrees)

Shell 1 BA 0 0 BB 120 60 Be 240 60 Shell 2 BD 52 52 BE 142 38 BF 232 52 BG 322 38 Shell 3 BH 109 71 BJ 229 49 BK 349 11 Table 5-14 Calculated Maximum Vessel Shell 1 Weld Exposures at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures.

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5-32

Vessel and Capsule Fluence Results Non-Proprietary Table 5-15 Calculated Maximum Vessel Shell 2 Weld Exposures at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures.

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Table 5-16 Calculated Maximum Vessel Shell 3 Weld Exposures at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures.

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5-33

Vessel and Capsule Fluence Results Non-Proprietary Table 5-17Calculated Maximum Vessel Circumferential Weld Exposures at the End of Cycle 21 (28.088 EFPY) and Projected to Future Exposures.

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Table 5-18 GGNS Calculated Vessel Fluence and Fluence Determined using dpa Attenuation.

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5-34

Vessel and Capsule Fluence Results Non-Proprietary Table 5-19 GGNS RPV Nozzle Locations.

Fluence Fluence Location Location Height Height Height Above Height Maximum Specified in Specified in Nozzle Vessel To Center Nozzle Reference Reference Center Zero to of Nozzle Nozzle Fluence Point [5-9] [5-9] Angles in Center of Relative Weld Relative to Relative to Relative to First Nozzle Nozzle toBAF on BAF Vessel '0' BAF Quadrant (inches) (inches) (inches) (inches) (inches) (inches)

Nl 172.31 -44.00 24.750 -31.62 199.7 -16.61 0° 26°15',

N2 179.31 -37.00 13.094 -30.45 197.0 -19.31 51°45',

7T)15' N12 366.00 149.69 1.880 148.75 364.7 148.39 15° N6 419.00 202.69 13.250 196.07 -a -a 39° a The desired fluence locations for nozzle N6 were not specified in Reference [5-9].

Table 5-20 GGNS RPV Nozzle Maximum Fluence Values.

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5-35

Vessel and Capsule Fluence Results Non-Proprietary Table 5-21GGNS RPV Nozzle Region Maximum Fluence Values in the First Quadrant at the Axial Locations Specified in Reference [5-9].

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Table 5-22GGNSRPV Nozzle Region Fluence Values at the Axial and Azimuthal Locations Specified in Reference [5-9].

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5-36

Vessel and Capsule Fluence Results Non-Proprietary Table 5-23 GGNS RPV Nozzle N6 Maximum FluenceValues With Margin Term Applied to Account for Above Core Fluence Uncertainty.

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5-37

Vessel and Capsule Fluence Results Non-Proprietary Table 5-24 Fluence Locations Specified by GEH for RPV Internals Mechanical Evaluation.

Core Support Radius from RPV Elevation from Elevation Above Structure/ Center Vessel Zero (BAF Component (inches) (inches) (inches)

Shroud" 106.4 Variable axial span Variable axial span Shroud Support Cylinder" 106.625 150.50 (Top) -65.81 Shroud Support Plate" Variable radial span 140.50 (Top) -75.81 Shroud Support Legs" 102.62 136.25 (Top) -80.06 Core PlateC Variable radial span 207.38 (Top) -8.93 Top Guide" Variable radial span 373.95 (Bottom) 157.64 211.38 (Top)

Core Plate Bolts" 102.88 -30.93 to -4.93 185.38 (Bottom)

Core Plate Bolt Nuts" 102.12 209.38 (Top) -6.93 Core Plate Wedges" 103.50 219.21 (Top) 2.9 382.63 (Top)

Top Guide Bolts" 105.94 153.57 to 166.32 369.88 (Bottom)

Top Guide Bolt Nuts" 105.06 378.38 (Bottom) 162.07 Top Guide Bolt Pinsd 106.31 373.38 (Bottom) 157.07 Control Rod Drive Coincident with RPV 51.00 (Top) -165.31 Housing! Centerline Coincident with RPV Control Rod Guide Tube f 208.133 (Top) -8.177 Centerline Coincident with RPV Orificed Fuel Support' 211.203 (Top) -5.107 Centerline 305.315 (Top)

Jet Pump Riser Brace" 108.88 83.995 to 89.005 300.305 (Bottom) 349.97 (Top)

Jet Pump Beam Bolte 115.03 130.73 to 133.66 347.04 (Bottom) 330.01 (Top)

Jet Pump Riser" 108.82 -58.435 to 113.7 157.875 (Bottom) 260.41 (Top),

Jet Pump Diffuser 107.63 -75.8 1 to 44. 1 40.50 (Bottom) a Variable axial span here indicates findmg the peak fluence over the entire surface of the shroud.

b Components with a radius and an axial elevation specified indicates that the peak must be found by sweeping over all azimuthal angles.

C For cylindrical components and plates, the peak fluence anywhere on the plane defining the plate has been reported.

d For components with discrete angular positions, the peak reported is the peak over the 360 degrees at the elevation and radius given.

e For these components, GEH has requested the peak as previously defined (over 360 degrees and not necessarily at the component), as well as the average over the axial extent given in the Table 5-24 at the peak angle location.

f At these locations, the value at the core centerline, and at the peak on the plane, are reported.

5-38

Vessel and Capsule Fluence Results Non-Proprietary Table 5-25 Fluence Evaluation at the End of Cycle 21 (28.088 EFPY) for use in RPV Internals Mechanical Evaluation.

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5-39

Vessel and Capsule Fluence Results Non-Proprietary Table 5-26 Fluence Evaluation at the 35 EFPY for use in RPV Internals Mechanical Evaluation.

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5-40

Vessel and Capsule Fluence Results Non-Proprietary Table 5-27 Fluence Evaluation at the 54 EFPY for use in RPV Internals Mechanical Evaluation.

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5-41

Vessel and Capsule Fluence Results Non-Proprietary

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Figure 5-1Azimuthal Variation of Maximum Vessel Fluence (E > 1.0 MeV) at the End of Cycle 21 (28.088 EFPY).

5-42

Vessel and Capsule Fluence Results Non-Proprietary

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Figure 5-2 Axial Variation of Peak Fluence (E > 1.0 MeV) in the Vessel at the End of Cycle 21 (28.088 EFPY).

5-43

Vessel and Capsule Fluence Results Non-Proprietary

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Figure 5-3 GGNS Vessel Roll Out Drawing Showing Beltline Weld Seam and Plate Locations. The Increase in the Axial Extent of the Beltline will Eventually Require that Radiation Damage Effects for Shell 3 Materials be Included in the PT Curve Analysis.

5-44

Vessel and Capsule Fluence Results Non-Proprietary

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Figure 5-4Axial Variation of Peak Fluence (E > 1.0 MeV) in the Vessel at 54 EFPY.

5-45

tvon-rroprtetary UNCERTAINTY ANALYSIS A comprehensive report has been prepared which documents the benchmarking of MPM transport methods for BWR plant analyses [6-1]. For each plant analyzed, a detailed uncertainty analysis was performed to estimate each source of uncertainty in the calculated fluence values.

This analysis made use of defined uncertainties and tolerances where possible, but some of the uncertainty estimates had to be based on estimates derived from data variation, such as the detailed power distribution and void fraction variations within a single cycle. This section of the report summarizes the uncertainty analysis performed for GGNS.

The geometry uncertainty assignments are from plant drawings referenced in Tables 2-2 through 2-4, and from generic assessments for cases where dimensional uncertainties are not available. Discussion of each uncertainty assumption is given below. Based on these uncertainty values, detailed evaluations were performed for the shroud, surveillance capsule, reactor pressure vessel, and N6 and N12 vessel nozzles. The uncertainty assessments for TORT results at the shroud, capsule, and vessel over the active fuel elevations are summarized in Table 6-1. The uncertainty assessments for 2D SYnthesis results at the shroud, capsule, and vessel over the active fuel regions are summarized in Table 6-2. The differences in the uncertainties between TORT and 2D SYnthesis are in the methods uncertainty as discussed further below.

Uncertainty estimates for the shroud and top guide welds are summarized in Tables 6-3 through 6-5. Finally, the vessel nozzle uncertainties are given in Table 6-6.

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6.1 Uncertainty Assumptions

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6-1

Uncertainty Analysis Non-Proprietary

[[

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6-2

Uncertainty Analysis Non-Proprietary

[[

]]

6-3

Uncertainty Analysis Non-Proprietary

[[

]]

6-4

Uncertainty Analysis Non-Proprietary 6.2 20/30 Uncertainty Evaluation over Active Fuel length The results for the uncertainty evaluation over the active fuel length are summarized in Tables6-1 and 6-2, based on the assumptions just discussed. These tables are applicable to the shroud, vessel, and surveillance capsule over the extent of the active fuel [[

]] As previously mentioned, a total uncertainty was derived by combining the independent individual contributors in quadrature.

3D TORT Uncertainty

[[

]]

6-5

Uncertainty Analysis Non-Proprietary

[[

]]

6.3 3D Model Uncertainty Evaluation for Shroud, Top Guide Weld, and Nozzle locations

[[

]]

6-6

Uncertainty Analysis Non-Proprietary

[[

]]

6-7

Uncertainty Analysis Non-Proprietary 6.4 Uncertainty Conclusions The detailed uncertainty analysis demonstrates that the MPM calculational methods provide fluence results within allowable tolerance bounds (+20%) for the reactor vessel, shroud welds, and surveillance capsules which lie within the axial active fuel region. This satisfies the requirements of RG 1.190. However, RG 1.190 does not address benchmarking outside this region. In the uncertainty analysis, evaluations of structures outside the beltline region have been made, and, especially above the core, [[

]]

Regulatory Guide 1.190 requires that the overall fluence calculation bias and uncertainty must be determined by an appropriate combination of the analytic uncertainty analysis results and the results of the uncertainty analysis based on the comparisons to theoperating reactor and simulator benchmark measurements. The regulatory guide states that this combination may be a weighted average that accounts for the reliability of the individualestimates. The regulatory guide goes on to state that if the analytical uncertainty at the 1 sigma level is greater than 30%,

the methodology of the regulatory guide is not applicable and theapplication will be reviewed on an individual basis. [[

]]

6.4 Section6 References

[6-1] [[

]]

[6-2] [[

]]

[6-3] McElroy, W.N., Ed., "LWR-PV-SDIP: PCA Experiments and Blind Test", NUREG/CR-1861,1981.

[6-4] [[

]]

[6-5] Maerker, R.E., et.al., "Application ofLEPRICON Methodology to LWR Pressure Vessel Dosimetry", Reactor Dosimetry: Methods, Applications, and Standardization, ASTM STP 1001, 1989,pp405-414.

[6-6] [[

]]

6-8

Uncertainty Analysis Non-Proprietary

[6-7] [[

]]

[6-8] [[

]]

[6-9] Fero, A.H., "Neutron and Gamma Ray Flux Calculations for the Venus PWR Engineering Mockup," NUREG/CR-4827 (WCAP-l1173), January, 1987.

[6-10] Tsukiyama, T., et. al., "Benchmark Validation of TORT Code Using KKM Measurement and its Application to 800 MWe BWR," Proceedings of the lIth International Symposium on Reactor Dosimetry, Brussels, Belgium, August 18-23, 2002, World Scientific, January 1, 2003.

[6-11] [[

]]

6-9

Uncertainty Analysis Non-Proprietary Table 6-1 Grand Gulf Shroud, Capsule, and Vessel Active Fuel Region TORT Calculational Fluence Uncertainty.

[[

]]

6-10

Uncertainty Analysis Non-Proprietary Table 6-2 Grand Gulf Shroud, Capsule, and Vessel Active Fuel Region 2D Synthesis Calculational Fluence Uncertainty.

[[

]]

6-11

Uncertainty Analysis Non-Proprietary Table 6-3 Estimated Maximum Uncertainty for Shroud and Top Guide Horizontal Welds.

Maximum Fluence Uncertainty Horizontal Weld (Percent)

HI [[ ]]

H2 [[ ]]

H3 [[ ]]

H4 [[ ]]

H5 [[ ]]

H6A [[ ]]

H6B [[ ]]

H7 [[ ]]

Table 6-4 Estimated Maximum Uncertainty for Shroud and Top Guide Vertical Welds.

Maximum Fluence Uncertainty Vertical Weld (Percent)

VI to V4 [[ ]]

V5, V6 [[ ]]

V7,V8 [[ ]]

V9 to V12 [[ ]]

V13,V14 [[ ]]

V15,V16 [[ ]]

V17,V18 [[ ]]

V19 to V22 [[ ]]

V23, V24 [[ ]]

6-12

Uncertainty Analysis Non-Proprietary Table 6-5 Estimated Maximum Weld IR Fluence Uncertainty for Shroud and Top Guide Vertical Welds.

Maximum Uncertainty to Weld IR Fluence Vertical Weld (Percent)

VI to V4 [[ ]]

V5, V6 [[ ]]

V7,V8 [[ ]]

V9 to V12 [[ ]]

V13,V14 [[ ]]

V15,V16 [[ ]]

V17,V18 [[ ]]

V19 to V22 [[ ]]

V23, V24 [[ ]]

Table 6-6 Estimated Maximum Uncertainty forVessel Nozzles N6 and NI2.

Maximum Fluence Uncertainty Nozzle Identification (Percent)

N6 [[ ]]

N12 [[ ]]

6-13

tvon-rroprtetary 7

SUMMARY

AND CONCLUSIONS A detailed three-dimensional transport calculation has been completed for GGNS encompassing operation over the first 20 fuel cycles [[ ]] This work was undertaken to calculate the best estimate neutron fluence, and its uncertainty, to the GGNS reactor pressure vessel, surveillance capsule, core shroud/top guide horizontal and vertical welds, welds,as as well as to several below core structures and nozzles. The calculations in the active fuel region were carried out using a three dimensional neutron transport calculation [[

]] Based on the calculations and analyses performed, the following conclusions have been made:

.. The transport calculations meet all of the criteria of RG 1.190 for evaluation of fluence to the shroud, surveillance capsule, and reactor vessel within the reactor beltline region.

The calculational methodology has been benchmarked to reactor mock-ups, calculational benchmarks, and measurements in other BWR plants. In addition, the calculations have been benchmarked against measurements in GGNS. Comparisons with dosimetry measurements at the GGNS surveillance capsule location at the end of cycle 1 were made and excellent agreement was found. [[

]]

• Fluence results have been obtained for all GGNS shroud and top guide welds. [[

]]

7-1

Summary and Conclusions Non-Proprietary

• Fluence results have been obtained for the reactor vessel and projected for future operation up to 54 EFPY. [[

]]

• [[

]]

• [[

]]

• Fluence calculations for the surveillance capsules indicated that the capsule lead factor is

[[ ]] Although the lead factor is less than unity, this is not a concern for GGNS since the BWRVIP surveillance program does not require any future capsule testing for GGNS.

• [[

]]

7-2

ivon-rroprtetary NOMENCLATURE ASTM American Society for Testing and Materials BAF bottom. of active fuel BWR boiling water reactor BWRVIP BWR Vessel Internals Project CIM calculated-to-measured ratio D dimension dpa displacements per atom EBZ enriched bottom zone EFPS effective full power seconds EFPY effective full power years ENDF evaluated nuclear data file EOC end-of-cycle GEH General Electric Hitachi GGNS Grand Gulf Nuclear Station ID inner diameter IGSCC irradiation assisted stress corrosion cracking IR inner radius LHGR linear heat generation rate LWR light water reactor MWd/MTU megawatt days per metric ton of uranium MOC middle-of-cycle MPM MP Machinery and Testing, LLC NMP-1 Nine Mile Point Unit 1 NMP-2 Nine Mile Point Unit 2 NRC U. S. Nuclear Regulatory Commission OD outer diameter OR outer radius ORNL Oak Ridge National Laboratory PCA pool critical assembly PT Pressure-Temperature PWR pressurized water reactor RBS River Bend Station RG Regulatory Guide RPV Reactor Pressure Vessel RSICC Radiation Safety Information Computational Center RTNDT Reference temperature for nil ductility transition RTpTS Reference temperature for pressurized thermal shock SE sequence exchange T vessel wall thickness 8-1

Nomenclature Non-Proprietary TAF top of active fuel

[[ ]]

8-2

ivon-rroprtetary SHROUD/TOP GUIDE WELD FLUENCE RESULTS AT THE END OF CYCLE 21 (28.088 EFPY EXPOSURE)

This appendix contains calculated fast fluence values (fluence for neutrons with energy above 1 MeV) for welds in the shroud and top guide. Fluence values for each weld are given at the IR, OR, and at positions 1/4, 1/2, and 3/4 of the distance between the IR and OR for an operation time of28.088 EFPY (the calculated end of cycle 21). Values are tabulated versus azimuthal angle for horizontal welds, and versus height above BAF for vertical welds.

[[

]]

A-I

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-IFast Fluence at Locations in the Top Guide for Weld HI vs.

Azimuth.

[[

]]

A-2

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-2Fast Fluence at Locations in the Top Guide for Weld H2 vs.

Azimuth.

[[

]]

A-3

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-3 Fast Fluence at Locations in the Shroud for Weld H3 vs.

Azimuth.

[[

]]

A-4

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-4 Fast Fluence at Locations in the Shroud for Weld H4 vs.

Azimuth.

[[

]]

A-5

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-5 Fast Fluence at Locations in the Shroud for Weld H5 vs.

Azimuth".

[[

]]

A-6

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-6 Fast Fluence at Locations in the Shroud for Weld H6A vs.

Azimuth".

[[

]]

A-7

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-7Fast Fluence at Locations in the Shroud for Weld H6B vs.

Azimuth.

[[

]]

A-8

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

JVon-]Jroprieta~

Appendix Table A-8Fast Fluence at Locations in the Shroud for Weld H7 vs.

Azimuth.

[[

]]

A-9

Shroud/Top Guide Weld Fluence Results at the End ofCyc1e 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-9 Fast Fluence at Locations in the Top Guide for Welds VI and V3 vs. Height above BAF.

[[

]]

Appendix Table A-IO Fast Fluence at Locations in the Top Guide for Welds V2 and V4 vs. Height above BAF.

[[

]]

A-IO

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

JVon-j>roprieta~

Appendix Table A-IIFast Fluence at Locations in the Top Guide for Weld V5 and V6 vs. Height above BAF.

[[

]]

A-II

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-12Fast Fluence at Locations in the Top Guide for Welds V7 and V8 vs. Radial Location.

[[

]]

A-12

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-13 Fast Fluence at Locations in the Top Guide for Welds V9 and VII vs. Height above BAF.

[[

]]

Appendix Table A-14Fast Fluence at Locations in the Top Guide for Welds VI0 and VI2 vs, Height above BAF.

[[

]]

A-I3

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-15Fast Fluence at Locations in the Shroud for Weld V13 and V14 vs. Height above BAF.

[[

]]

A-14

Shroud/Top Guide Weld Fluence Results at the End ofCyc1e 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-16 Fast Fluence at Locations in the Shroud for Weld VIS and V16 vs. Height above BAF.

[[

]]

A-I5

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-17Fast Fluence at Locations in the Shroud for Welds V17 and V18 vs. Height above BAF.

[[

]]

A-16

Shroud/Top Guide Weld Fluence Results at the End ofCyc1e 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-IS Fast Fluence at Locations in the Shroud for Welds V19 and V21 vs. Height above BAF.

[[

]]

Appendix Table A-19 Fast Fluence at Locations in the Shroud for Welds V20 and V22 vs. Height above lBAF.

[[

]]

A-I7

Shroud/Top Guide Weld Fluence Results at the End of Cycle 21 (28.088 EFPY Exposure)

Non-Proprietary Appendix Table A-20Fast Fluence at Locations in the Shroud for Welds V23 and V24 vs. Height above BAF.

[[

]]

A-I8

tvon-rropnetary SHROUD/TOP GUIDE WELD FLUENCE RESULTS AFTER 35 EFPY EXPOSURE*

This appendix contains calculated fast fluence values (fluence for neutrons with energy above 1 MeV) for welds in the shroud and top guide. Fluence values for each weld are given at the IR, OR, and at positions 1/4, 1/2, and 3/4 of the distance between the IR and OR for an exposure of35 EFPY (6.912 EFPY beyond the calculated end of cycle 21). Values are tabulated versus azimuthal angle for horizontal welds, and versus height above BAF for vertical welds.

[[

]]

B-1

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table B-IFast Fluence at Locations in the Top Guide for Weld HI vs.

Azimuth.

[[

]]

B-2

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table B-2 FastFluence at Locations in the Top Guide for Weld 82 vs.

Azimuth.

[[

]]

B-3

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table B-3Fast Fluence at Locations in the Shroud for Weld H3 vs.

Azimuth.

[[

]]

B-4

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure lVon-jDroprieta~

Appendix Table B-4Fast Fluence at Locations in the Shroud for WeldH4 vs.

Azimuth.

[[

]]

B-5

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table B-5Fast Fluence at Locations in the Shroud for Weld H5 vs.

Azimuth".

[[

]]

B-6

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary

\

Appendix Table B-6Fast Fluence at Locations in the Shroud for Weld H6A vs.

Azimuth",

[[

]]

B-7

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table B-7 Fast Fluence at Locations in the Shroud for Weld H6B vs.

Azimuth.

[[

]]

B-8

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table B-8Fast Fluence at Locations in the Shroud for, Weld H7 vs.

Azimuth.

[[

]]

B-9

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table B-9Fast Fluence at Locations in the Top Guide for Welds VI and V3 vs. Height above BAF.

[[

]]

Appendix Table B-I0 Fast Fluence at Locations in the Top Guide for Welds V2 and V4 vs. Height above BAF.

[[

]]

B-10

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure (\

Non-Proprietary Appendix Table 8-11 Fast Fluence at Locations in the Top Guide for Weld V5 and V6 vs. Height above 8AF.

[[

]]

B-11

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table 8-12 Fast Fluence at Locations in the Top Guide for Welds V7 and V8 vs. Radial Location.

[[

]]

".

B-12

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table B-13Fast Fluence at Locations in the Top Guide for Welds V9 and VII vs. Height above BAF.

[[

]]

Appendix Table B-14Fast Fluence at Locations in the Top Guide for Welds VI0 and V12 vs. Height above BAF.

[[

]]

B-13

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table B-15Fast Fluence at Locations in the Shroud for Weld VI3 and V14 vs. Height above BAF.

[[

]]

B-14

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table B-16 Fast Fluence at Locations in the Shroud for Weld VIS and V16 vs, Height above BAF.

[[

]]

B-15

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table 8-17 Fast Fluence at Locations in the Shroud for Welds V17 and VI8 vs. Height above BAF.

[[

]]

B-16

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table B-18Fast Fluence at Locations in the Shroud for Welds V19 and V21 vs. Height above BAF. .

[[

]]

Appendix Table B-19FasfFluence at Locations in the Shroud for Welds V20 and V22 vs. Height above BAF.

[[

]]

B-17

Shroud/top guide Weld Fluence Results After 35 EFPY Exposure Non-Proprietary Appendix Table 8-20 Fast Fluence at Locations in the Shroud for Welds V23 and V24 vs. Height above BAF.

[[

]]

B-18

ivon-rroprtetary SHROUD/TOP GUIDE WELD FLUENCE RESULTS AFTER 54 EFPY EXPOSURE This appendix contains calculated fast fluence values (fluence for neutrons with energy above 1 MeV) for welds in the shroud and top guide. Fluence values for each weld are given at the IR, OR, and at positions 1/4, 1/2, and 3/4 of the distance between the IR and OR for an exposure of 54 EFPY (25.912 EFPY beyond the calculated end of cycle 21). Values are tabulated versus azimuthal angle for horizontal welds, and versus height above BAF for vertical welds.

[[

]]

C-l

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix TableC-1 Fast Fluence at Locations in the Top Guide for Weld HI vs.

Azimuth.

[[

]]

C-2

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-2Fast Fluence at Locations in the Top Guide for Weld H2 vs.

Azimuth.

[[

]]

C-3

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-3Fast Fluence at Locations in the Shroud for Weld H3 vs.

Azimuth.

[[

]]

C-4

Shroud/top guide We1d Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-4Fast Fluence at Locations in the Shroud for Weld H4 vs.

Azimuth.

[[

]]

C-5

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-5 Fast Fluence at Locations in the Shroud for Weld H5 vs.

Azimuth",

[[

]]

C-6

Shroud/top guide We1d Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-6 Fast Fluence at Locations in the Shroud for Weld H6A vs.

Azimuth",

[[

]]

C-7

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-7Fast Fluence at Locations in the Shroud for Weld H6B vs.

Azimuth.

[[

]]

C-8

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure lVon-jDroprieta~

Appendix Table C-8 Fast Fluence at Locations in the Shroud for Weld H7 vs.

Azimuth.

[[

]]

C-9

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-9 Fast Fluence at Locations in the Top Guide for Welds VI and V3 vs. Height above BAF.

[[

]]

Appendix Table C-IO Fast Fluence at Locations in the Top Guide 'for Welds V2 and V4 vs. Height above BAF.

[[

]]

C-IO

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-lllFast Fluence at Locations in the Top Guide for Weld V5 and V6 vs. Height above BAF.

[[

]]

C-ll

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-12 Fast Fluence at Locations in the Top Guide for Welds V7 and V8 vs. Radial Location.

[[

]]

C-12

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-13 Fast Fluence at Locations in the Top Guide for Welds V9 and VII vs. Height above BAF.

[[

]]

Appendix Table C-14 Fast Fluence at Locations in the Top Guide for Welds VIO and VI2 vs. Height above BAF.

[[

]]

C-13

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-15Fast Fluence at Locations in the Shroud for Weld V13 and V14 vs. Height above BAF.

[[

]]

C-14

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-16Fast Fluence at Locations in the Shroud for Weld VI5 and V16 vs. Height above BAF.

[[

]]

C-15

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-17 Fast Fluence at Locations in the Shroud for Welds V17 and VIS vs. Height above BAF.

[[

]]

C-16

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-18 Fast Fluence at Locations in the Shroud for Welds V19 and V21 vs. Height above BAF.

[[

]]

Appendix Table C-19Fast Fluence at Locations in the Shroud for Welds V20 and V22 vs. Height above BAF.

[[

]]

C-17

Shroud/top guide Weld Fluence Results After 54 EFPY Exposure Non-Proprietary Appendix Table C-20 Fast Fluence at Locations in the Shroud for Welds V23 and V24 vs. Height above BAF.

[[

]]

C-18

Attachment 4 to GNRO-2016/00034 NEDO-33669, Revision 1, April 2015, Non-Proprietary Version, "Grand Gulf Nuclear Station Fluence Effect on RPV Internal Components At EPU Operating Conditions", (Non-Proprietary Version of Attachment 1 to Engineering Report GGNS-NE-15-00003, Revision 0)

HITACHI GE Hitachi Nuclear Energy NEDO-33669 Revision 1 April 2015 Non-Proprietary Information - Class I (Public)

Grand Gulf Nuclear Station Fluence Effect on RPV Internal Components at EPU Operating Conditions Copyright 2015 GE-Hitachi Nuclear Energy Americas, LLC All Rights Reserved

NEDO-33669 Revision I Non-Proprietary Information - Class I (Public)

INFORMATION NOTICE This is a non-proprietary version of the document NEDC-33669P, Revision 1, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here [[ ]].

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document is furnished for the purposes of supporting the Entergy license amendment request for a License Renewal at Grand Gulf Nuclear Station in proceedings before the U.S. Nuclear Regulatory Commission. The only undertakings of GEH with respect to information in this document are contained in the contracts between GEH and its customers or participating utilities, and nothing contained in this document shall be construed as changing that contract. The use of this information by anyone for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, GEH makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.

II

NEDO-33669 Revision 1 Non-Proprietary Information - Class I (Public)

REVISION

SUMMARY

o Initial Issue 1 Evaluation revised using the fluence provided by Entergy in Reference 2. The GEH calculated fluence was used in Revision o.

III

NEDO-33669 Revision 1 Non-Proprietary Information - Class 1(Public)

TABLE OF CONTENTS Page 1 BACKGROUND AND SCOPE OF WORK 1 1.1 Background 1 1.2 Scope of Work 1 2 REFERENCES '" 2 3 EVALUATION 3 3.1 Acceptance Criteria 3 3.2 Material Properties of Irradiated Type 304 Stainless Steel. .4 3.3 Neutron Fluence 4 3.4 Analysis Inputs 6 3.5 Results 6 3.6 Supporting Evaluation 8 4 CONCLUSIONS 10 IV

NEDO-33669 Revision 1 Non-Proprietary Information - Class I (Public) 1 BACKGROUND AND SCOPE OF WORK 1.1 Background Entergy intends to apply for the license renewal of Grand Gulf Nuclear Station for 20 additional years. The information on the fluence effect on the Reactor Pressure Vessel (RPV) Internal Core Support Structure (CSS) components at Extended Power Uprate (EPU) operating conditions (Reference 1) is needed to support the evaluations for the license renewal. Such evaluation was performed previously with the GEH calculated fluence in Revision 0 of this report. In this revision, the structural evaluation is updated with fluence inputs from Entergy.

1.2 Scope of Work The fluence effect on RPV internal components at EPU operating conditions is to be evaluated per the project contract order. Only the RPV internal CSS components need to be evaluated.

The fluence levels for a 60-year plant life at EPU operating conditions with 100/0 margin are considered. The structural acceptance criteria for RPV internal CSS components are evaluated.

All CSS component stresses/forces are calculated for the fluence levels and compared with the acceptance criteria for a 60-year plant life at EPU operating conditions.

The RPV Internal CSS components are:

• Shroud Support

• Shroud (Including Welds)

• Core Plate (Including Core Plate Bolts)

• Top Guide (Including Top Guide Bolts)

• Control Rod Drive Housing (CRDH)

• Control Rod Guide Tube (CRGT)

• Orificed Fuel Support (OFS)

• Peripheral Fuel Support (PFS)

The non-CSS component of the Jet Pump Beam Bolt is also evaluated for the fluence effect.

NEDO-33669 Revision 1 Non-Proprietary Information - Class I (Public) 2 REFERENCES

~~i~~in'(:

GE Hitachi Nuclear Energy, "Safety Analysis Report for Grand Gulf Nuclear Station Constant Pressure Power Uprate," NEDC-33477P, Revision 0, August 2010.

2 MP Machinery and Testing, LLC, "Neutron Transport Analysis for Grand Gulf Nuclear Station," MPM Report Number MPM-814779 Revision 3, February 2015.

3 "BWR Core Plate Inspection and Flaw Evaluation Guidelines," BWRVIP-25, December 1996.

4 ASME Boiler & Pressure Vessel Code,Section III, Division 1 - Subsection NG, 2007 Edition.

2

NEDO-33669 Revision 1 Non-Proprietary Information - Class I (Public) 3 EVALUATION 3.1 Acceptance Criteria The RPV internal CSS components need to comply with the criteria for irradiation. For austenitic stainless steel components subjected to a lifetime neutron fluence level less than 1E21 nvt (Er > 1 MeV), the base material requires no special consideration in addition to meeting basic ASME Code requirements (Reference 4). For austenitic stainless steel components with less than 5E20 nvt (Er > 1 MeV), the weld material requires no special consideration, where nvt is fluence, Ej is Energy, and MeV is Million electron volts.

At fluence greater than 1E21 nvt (E, > 1 MeV) for 304 or 304L base material and 5E20 nvt (Er > 1 MeV) for 308 or 308L welds, that portion of the component and weld exposed to this greater fluence shall meet the following criteria in addition to American Society of Mechanical Engineers (ASME) code requirements (Reference 4):

a. For normal and upset conditions:

and  ::; 0.45 cunif

b. For emergency conditions:

Pm+Qm < 1

-E- - zcunif and s 0.67 cunif

c. For faulted conditions:

and  ::; 0.90 cunif where:

Pm: primary membrane stress Pb: primary bending stress Qm: secondary membrane stress Q: secondary stress (including membrane and bending stresses)

E: Young's modulus Cunif: designates the uniform elongation which has a value of 0.5 percent for these conditions at 550°F and has a value of 0.6 percent for these conditions at 75°F. Between 75°F and 550°F, the value of cunif may be linearly interpolated between 0.5 and 0.6 percent.

308 or 308L welds subjected to fluences greater than 5E20 nvt (Ef> 1 MeV) shall either be limited to a maximum stress of 5,000 psi resulting from any operating and accident load or be full penetration welds with a minimum quality of 0.9 as determined by Table NG-3352-1 of ASME Code (Reference 4).

3

NEDO-33669 Revision 1 Non-Proprietary Information - Class I (Public)

In addition to meeting ASME code threaded structural fastener requirements, the bolted joint shall be designed to provide for relaxation (thermal and irradiation) using the end of plant life fluence values.

3.2 Material Properties of Irradiated Type 304 Stainless Steel The applicable material properties for this evaluation are the stress relaxation curve due to neutron irradiation and the material ductility degradations as the fluence level exceeds 5E20 nvt (Ef> 1 MeV).

The ultimate tensile strength increases as fluence level increases. The yield strength increases as fluence level increases. The proportional elastic limit increases as fluence level increases.

Uniform elongation decreases rapidly between 5E20 nvt and lE21 nvt (Er > IMeV). Above lE21 nvt (Er > IMeV), the materials ductility does not degrade further. Stress relaxes as fluence level increases. The stress relaxation reduces the preload value in structural bolts.

The elastic modulus and uniform elongation for Type 304 Stainless Steel are listed in Table 1.

Table 1 Type 304 Stainless Steel Properties Elastic Modulus, xl 06 psi 28.3 25.6 Uniform Elongation, &lll1if

&lll1ij 0.006 0.005 1/3&unif 1/3&unij 0.0020 0.0017 0.45&1lI1if 0.45&1lI1ij 0.0027 0.0023 1/2&1lI1if 1/2&1lI1ij 0.0030 0.0025 0.67 &unif

&unij 0.0040 0.0034 2/3&llnif 2/3&llnij 0.0040 0.0033 0.9&1lI1if 0.9&1lI1ij 0.0054 0.0045 3.3 Neutron Fluence Neutron fluence at 54 Effective Full Power Years (EFPY) (equivalent to a 60 year plant life at a 90% capacity factor) is provided by Entergy in Reference 2.

Per Entergy's request, the provided total 54 EFPY fluences are multiplied by 110%, and listed in Column 4 of Table 2. The 110% fluences at 54 EFPY are used for the evaluation of the fluence effect on RPV internal CSS components.

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Table 2 Peak Neutron Fluence 5

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3.4 Analysis Inputs Fluence inputs are provided by Entergy in Reference 2. Other inputs used for this analysis are based on the Grand Gulf Nuclear Station EPU RPV internal mechanical analysis (Reference 1).

Grand Gulf Nuclear Station has decided to install a replacement Steam Dryer. The new replacement Steam Dryer has no effect on the existing plant seismic/dynamic loads and Reactor Internal Pressure Differentials (RIPDs). All other design inputs remain unchanged from Reference 1 and are applicable for this analysis.

3.5 Results The fluence effect on RPV internal CSS components at EPU operating conditions is evaluated.

The non-CSS component of the Jet Pump Beam Bolt is also evaluated for the fluence effect. The key results affected by fluence are listed in Table 3.

Table 3 Key Results Affected by Fluence 6

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3.6 Supporting Evaluation 8

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4 CONCLUSIONS GEH evaluated the effect of fluence on RPV internal components at EPU operating conditions for a 60-year plant life (54 EFPY). The fluence values at 54 EFPY with additional 10% margins were used for the irradiation compliance analysis of RPV internal components at EPU operating conditions.

The RPV internal CSS components meet the irradiation criteria at EPU operating conditions for the 110% fluence levels at 54 EFPY. The jet pump beam bolts also meet the criteria.

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