L-20-274, Amendment Request to Update Analytical Methods Used to Develop Reactor Coolant System Pressure and Temperature Limits

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Amendment Request to Update Analytical Methods Used to Develop Reactor Coolant System Pressure and Temperature Limits
ML20304A215
Person / Time
Site: Beaver Valley
Issue date: 10/30/2020
From: Penfield R
Energy Harbor Nuclear Corp
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-20-274
Download: ML20304A215 (34)
v • d • e


Text

energy harbor Energy Harbor Nuclear Corp.

Beaver Valley Power Station P.O. Box 4 Shippingport, PA 15077 Rod L. Penfield 724-682-5234 Site Vice President, Beaver Valley Nuclear October 30, 2020 10 CFR 50.90 L-20-274 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit Nos. 1 and 2 BVPS-1 Docket No. 50-334, License No. DPR-66 BVPS-2 Docket No. 50-412, License No. NPF-73 Amendment Request to Update Analytical Methods Used to Develop Reactor Coolant System Pressure and Temperature Limits Pursuant to 10 CFR 50.90, Energy Harbor Nuclear Corp. hereby requests an amendment to Technical Specification 5.6.4, "Reactor Coolant System (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," Item b, to replace the currently referenced analytical methods with more recent analytical methods found acceptable by the Nuclear Regulatory Commission (NRC) staff for calculating reactor vessel neutron fluence and reactor coolant system pressure and temperature limits when updating the reactor coolant system Pressure and Temperature Limits Report.

Energy Harbor Nuclear Corp.'s evaluation of the proposed change is provided in Enclosure A. The technical justification for using the proposed methodology in the extended beltline region for Beaver Valley Power Station, Unit No. 1 (BVPS-1) and Unit No. 2 (BVPS-2) is documented in Enclosure B.

Approval of the proposed amendment is requested by November 1, 2021. The amendment shall be implemented within 30 days of approval.

As a result of reactor vessel Surveillance Capsule Y test data, Energy Harbor Nuclear Corp. plans to update the pressure and temperature limits for BVPS-2 in accordance with the proposed requirements of Technical Specification 5.6.4. The revised pressure and temperature limits would be implemented at BVPS-2, prior to the expiration of the

Beaver Valley Power Station, Unit No. 2 L-20-274 Page 2 current pressure and temperature limits at 30 effective full power years of operation (that is, prior to approximately March of 2022).

There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Manager, Fleet Licensing, at (330) 696-7208.

I declare under penalty of perjury that the foregoing is true and correct. Executed on October _]Q___, 2020.

Sincerely, Pc-nlield, Rod 55166 site vp Penfield, Rod 55166 ~:~n37/'i~~~i;g4~il:cumcnt Docu ~ .

Rod L. Penfield

Enclosures:

A. Evaluation of Proposed Amendment B. Justification of Using RAPTOR-M3G for Reactor Pressure Vessel Extended Beltline Materials at Beaver Valley Units 1 and 2 cc: NRC Region I Administrator NRC Resident Inspector NRR Project Manager Director BRP/DEP Site BRP/DEP Representative

Enclosure A L-20-274 Evaluation of Proposed Amendment (14 pages follow)

Evaluation of Proposed Amendment Page 1 of 11

Subject:

Proposed Revision of Analytical Methods Specified in Technical Specification 5.6.4, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," for Beaver Valley Power Station, Unit Nos. 1 and 2 TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specification Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed change

3.0 TECHNICAL EVALUATION

3.1 WCAP-14040-A Methodology 3.2 WCAP-18124-NP-A Methodology 3.3 Current Pressure and Temperature Limit Report Values 3.4 Conclusion

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements / Criteria 4.2 Significant Hazards Consideration 4.3 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

S ATTACHMENT

1. Technical Specification Page Markups

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 2 of 11 1.0

SUMMARY

DESCRIPTION The proposed amendment would revise Technical Specification 5.6.4, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," to update the specified analytical methods used to determine reactor coolant system pressure and temperature limits for operation of the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2).

The proposed change to Technical Specification 5.6.4 is requested to permit use of more recent analytical methods found acceptable by the Nuclear Regulatory Commission (NRC) staff when calculating future reactor vessel neutron fluence and reactor coolant system pressure and temperature limits for updates to the BVPS-1 and BVPS-2 Pressure and Temperature Limits Report.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation Components of the reactor coolant system are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. Pressure and temperature changes are limited during reactor coolant system heatup and cooldown to be within the design assumptions and the stress limits for cyclic operation.

Pressure and temperature limit curves have been established for heatup, cooldown, and hydrostatic testing. Each pressure and temperature limit curve defines an acceptable region for normal operation. The curves are used during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. The minimum temperature required for inservice hydrostatic testing is also specified.

2.2 Current Technical Specification Requirements Technical Specification 5.6.4, Item b requires that the analytical methods used to determine the reactor coolant system pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

MB3319 and MB3320)," dated October 8, 2002.

  • [Westinghouse topical report] WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

The methodology listed in WCAP-14040-NP-A was used with two exceptions:

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 3 of 11

  • [American Society of Mechanical Engineers Boiler and Pressure Vessel Code] ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T [Pressure and Temperature] Limits for Section XI, Division 1."
  • ASME,Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 1996 version.

2.3 Reason for the Proposed Change The current BVPS-2 reactor coolant system pressure and temperature limit curves presented in the PTLR are applicable until 30 effective full power years (EFPY) of operation is reached.

In order to develop BVPS-2 pressure and temperature limit curves for operation beyond 30 EFPY, the methodology specified in Technical Specification 5.6.4, Item b, must be updated to permit use of more recent NRC staff-approved analytical methods. The more recent analytical methods are provided in two topical reports, WCAP-14040-A, Revision 4, and WCAP-18124-NP-A, Revision 0, "Fluence Determination with RAPTOR-M3G and FERRET," that have been found acceptable for use by the NRC.

Topical report WCAP-14040-A, Revision 4, provides a methodology to develop pressure and temperature limit curves, incorporates the two exceptions that are currently presented in Technical Specification 5.6.4, Item b, and provides a neutron fluence calculation methodology that adheres to the guidance in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." WCAP-18124-NP-A, Revision 0, provides an alternative neutron fluence calculation methodology that adheres to the guidance in Regulatory Guide 1.190.

2.4 Description of the Proposed Change The proposed license amendment would replace the methods currently specified in Technical Specification 5.6.4 for use in determining the reactor coolant system pressure and temperature limits with more recent NRC staff-approved analytical methods.

The proposed license amendment would revise Technical Specification 5.6.4 to:

  • Permit fluence calculations to be determined in accordance with the methodology described in topical report WCAP-14040-A, Revision 4, or topical report WCAP-18124-NP-A, Revision 0.

More specifically, the proposed amendment would change Item b of Technical Specification 5.6.4 to read:

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004. WCAP-18124-NP-A, Revision 0, "Fluence Determination with

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 4 of 11 RAPTOR-M3G and FERRET," July 2012, may be used as an alternative to Section 2.2 of WCAP-14040-A, Revision 4.

3.0 TECHNICAL EVALUATION

3.1 WCAP-14040-A Methodology The NRC has found the methodology contained in topical report WCAP-14040-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 3, acceptable for referencing in licensing applications subject to three conditions as described in the February 27, 2004 NRC final safety evaluation. Revision 3 of the topical report was reissued in May of 2004 as Revision 4 (Accession No. ML050120209) with the referenced NRC acceptance letter and safety evaluation incorporated. The three conditions and responses indicating how each condition has been addressed are presented below.

1. Licensees who wish to use WCAP-14040, Revision 3, as their PTLR methodology must provide additional information to address the methodology requirements discussed in Provision 2 in the table of Attachment 1 to Generic Letter 96-03 related to the reactor pressure vessel material surveillance program.

Response

The minimum methodology requirements associated with Provision 2 in the table of Attachment 1 to Generic Letter 96-03 require that licensees:

  • Briefly describe the surveillance program. Licensees should identify by title and number report containing the Reactor Vessel Surveillance Program and surveillance capsule reports.

BVPS-1 PTLR Section 5.2.2 and BVPS-2 PTLR Section 5.2.2, both titled "Reactor Vessel Material Surveillance Program," describe the methodology requirements discussed in Provision 2 in the table of Attachment 1 to Generic Letter 96-03 related to the reactor pressure vessel material surveillance program. Both programs are in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirements."

The most recent post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens for BVPS-1 was performed in accordance with 10 CFR 50, Appendix H and American Society for Testing and Materials (ASTM)

Specification E185-82 (WCAP-17896-NP, Revision 0, "Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," September 2014, Accession No. ML14288A393).

The most recent post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens for BVPS-2 was performed in accordance with 10 CFR 50, Appendix Hand ASTM Specification E185-82 (WCAP-18558-NP, Revision 0, "Analysis of Capsule Y from the Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," June 2020).

The NRC approved the current Reactor Vessel Surveillance Capsule Withdrawal Schedule for BVPS-1 in a letter dated July 2, 2018 (Accession No. ML18164A082).

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 5 of 11 The NRC approved the current Reactor Vessel Surveillance Capsule Withdrawal Schedule for BVPS-2 in a letter dated July 17, 2014 (Accession No. ML13242A266).

In a letter dated June 5, 2019 (Accession No. ML19126A195) the NRC staff approved an exemption from the testing and report submittal requirements of Appendix H to 10 CFR 50 for BVPS-2 reactor vessel surveillance capsule Y, that was removed from the reactor vessel on October 29, 2018. The exemption request stated that If a decision is made to operate BVPS-2 beyond October 31, 2021, a revised capsule testing schedule would be submitted for NRC approval prior to October 31, 2021. Surveillance Capsule Y testing and analysis began in September of 2019.

2. Contrary to the information in WCAP-14040, Revision 3, licensees use of the provisions of ASME Code Cases N-588, N-640, or N-641 in conjunction with the basic methodology contained in WCAP-14040, Revision 3, does not require an exemption since the provisions of these Code Cases are contained in the edition and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a. When published, the approved revision (Revision 4) of topical report WCAP-14040 should be modified to reflect this NRC staff conclusion.

Response

Table A-1, "Status of ASME Nuclear Code Cases Associated with the P-T Limit Curve/COMS Methodology," of Appendix A, "Relevant ASME Nuclear Code Cases" was updated in Revision 4 of WCAP-14040-A to indicate the date, edition and addenda of ASME Code,Section XI, when the referenced code cases were approved by the ASME. As indicated in Condition 2, the editions and addenda of ASME Code,Section XI, that are listed in Table A-1 have been incorporated by reference in 10 CFR 50.55a.

3. As stated in WCAP-14040, Revision 3, until Appendix G to 10 CFR Part 50 is revised to modify or eliminate the existing reactor pressure vessel flange minimum temperature requirements or an exemption request to modify or eliminate these requirements is approved by the NRC for a specific facility, the stated minimum temperature must be incorporated into a facility's pressure and temperature limit curves.

Response

This amendment proposes that Revision 4 of WCAP-14040-A be implemented for the PTLR methodology at BVPS-1 and BVPS-2. In accordance with Section 2.9 of WCAP-14040-A, Revision 4, the reactor pressure vessel flange minimum temperature requirement will continue to be incorporated into the BVPS-1 and BVPS-2 pressure and temperature limit curves, until Appendix G to 10 CFR Part 50 is revised to modify or eliminate the existing reactor pressure vessel flange requirements, or an exemption request to modify or eliminate these requirements is approved by the NRC.

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 6 of 11 3.2 WCAP-18124-NP-A Methodology The NRC staff has found the calculational fluence methodology described in WCAP-18124-NP-A, Revision 0, acceptable for use in calculating reactor pressure vessel neutron fluence subject to two conditions described in the June 15, 2018 NRC final safety evaluation (Accession No. ML18156A066). The two conditions and responses indicating how the conditions will be addressed are presented below.

1. Applicability of WCAP-18124-NP, Revision 0, is limited to the reactor pressure vessel region near the active height of the core based on the uncertainty analysis performed and the measurement data provided. Additional justification should be provided via additional benchmarking, fluence sensitivity analysis to response parameters of interest (for example, pressure-temperature limits, material stress and strain), margin assessment, or a combination thereof, for applications of the method to components including, but not limited to, the reactor pressure vessel upper circumferential weld, and reactor coolant system inlet and outlet nozzles and reactor vessel internal components.

Response

A plant-specific technical justification for using WCAP-18124-NP Revision 0 methodology in the extended beltline region has been performed for BVPS-1 and BVPS-2 and is documented in a report titled "Justification of Using RAPTOR-M3G for Reactor Pressure Vessel Extended Beltline Materials at Beaver Valley Units 1 and 2."

Based on the fluence uncertainty analysis, benchmarking, and margin assessment that are discussed in the report, it is concluded that WCAP-18124-NP Revision 0 methodology is acceptable to use when calculating fluence for the BVPS-1 and BVPS-2 extended beltline materials. For BVPS-1, there is significant margin between the pressurized thermal shock reference temperature values and adjusted reference temperature values of the limiting beltline region materials and the limiting extended beltline region materials. As a result, no bias factor will be applied to the BVPS-1 extended beltline region material fluence values. For BVPS-2, the report identifies that a bias factor of 1.10 must be applied to the calculated fast neutron fluence of the BVPS-2 limiting extended beltline region materials.

2. Least squares adjustment is acceptable if the adjustments to the measured or calculated ratios and to the calculated spectra values are within the assigned uncertainties of the calculated spectra, the dosimetry measured reaction rates, and the dosimetry reaction cross sections. Should this not be the case, the user should re-examine both measured and calculated values for possible errors. If errors cannot be found, the particular values causing the inconsistency should be disqualified.

Response

For BVPS-1 and BVPS-2, least squares adjustment results are typically reported for information only. If least squares adjustment is used to modify calculated fluence values for BVPS-1 and BVPS-2 during future reactor vessel integrity evaluations, the requirements of Limitation 2 will be met.

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 7 of 11 3.3 Current Pressure and Temperature Limit Report Values Except for the neutron fluence values, the current BVPS-1 PTLR values (Accession No. ML19105A881), and the current BVPS-2 PTLR values (Accession No. ML14133A107) were determined using methods consistent with Revision 2 of WCAP-14040-A and the plant specific allowances to use ASME Code Case N-640 and ASME Code Section XI, Appendix G (as permitted by Technical Specification 5.6.4 ).

The neutron fluence values supporting the current BVPS-1 PTLR, and the neutron fluence values supporting the current BVPS-2 PTLR, were determined using the updated methods described in Regulatory Guide 1.190.

The NRC previously reviewed the BVPS-1 reactor coolant system pressure and temperature limits developed with the updated fluence methodology when Technical Specification Amendment 249 was issued for BVPS-1 in 2002 (Accession No. ML020520701 ). The NRC previously reviewed BVPS-2 pressure and temperature limits developed with the updated fluence methodology during a review of the BVPS-2 PTLR performed in 2014 (Accession No. ML14251A558)."

3.4 Conclusion The proposed change in methodology for determining reactor coolant system pressure and temperature limits and calculating neutron fluence values is acceptable since the conditions and limitations identified by the NRC staff would be addressed as described above, and the methodologies have previously been found acceptable by the staff.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements / Criteria The proposed amendment has been evaluated to determine whether applicable regulations and requirements continue to be met as described below.

Appendix A to 10 CFR 50, General Design Criterion 14, "Reactor coolant pressure boundary," requires the design, fabrication, erection, and testing of the reactor coolant pressure boundary so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Appendix A to 10 CFR 50, General Design Criterion 30, "Quality of reactor coolant pressure boundary," requires, in part, that components comprising the reactor coolant pressure boundary be designed, fabricated, erected, and tested to the highest quality standards practical.

Appendix A to 10 CFR 50, General Design Criterion 31, "Fracture Prevention of reactor coolant pressure boundary," requires that:

The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 8 of 11 operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

10 CFR 50.36(c)(5), "Administrative controls," are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. The proposed amendment would continue to assure operation of the facility in a safe manner by updating the methodology used for determining reactor coolant system pressure and temperature limits and calculating neutron fluence values specified in the administrative controls section of Technical Specifications to more recent methodology found acceptable by the NRC staff.

10 CFR 50.60, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation," paragraph (a) states:

Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the certifications required under §50.82(a)(1) have been submitted, must meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in appendices G and H to this part.

The methodology used to determine the pressure and temperature limits in the PTLR and low temperature overpressure protection system enable temperature must comply with the specific requirements of Appendices G and H to 10 CFR 50.

Appendix G to 10 CFR 50, "Fracture Toughness Requirements," requires the establishment of pressure and temperature limits for specific material fracture toughness requirements of the reactor coolant pressure boundary materials.

Appendix H to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirements," establishes requirements related to facility reactor pressure vessel material surveillance programs.

In a February 27, 2004 safety evaluation, the NRC staff concluded that the methodology specified in WCAP-14040, Revision 3, addresses reactor pressure vessel minimum temperature requirements in a way which is consistent with Appendix G to 10 CFR 50 and Appendix G to Section XI of the ASME Code. The NRC Staff also concluded that the basic methodology specified in WCAP-14040, Revision 3, for establishing pressure and temperature limit curves meets the regulatory requirements of Appendix G to 10 CFR 50 and the guidance provided in Standard Review Plan, Section 5.3.2, "Pressure-Temperature Limits."

The BVPS-1 and BVPS-2 Reactor Vessel Material Surveillance programs comply with the requirements of Appendix H to 10 CFR 50. Implementation of the proposed amendment will not affect the compliance of the Reactor Vessel Material Surveillance program with Appendix H to 10 CFR 50.

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 9 of 11 To satisfy the requirements of both Appendix G to 10 CFR 50 and 10 CFR 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events," methods for determining the fast neutron fluence are necessary to estimate the fracture toughness of the pressure vessel materials. Regulatory Guide 1.190, "Calculational and Dosimetry Methods For Determining Pressure Vessel Neutron Fluence," dated March 2001, provides state-of-the-art calculations and measurement procedures that are acceptable to the NRC staff for determining pressure vessel fluence. The guide is intended to ensure the accuracy and reliability of the fluence determination required by General Design Criteria 14, 30, and 31 of Appendix A to 10 CFR 50.

The February 27, 2004 safety evaluation states that the NRC staff has reviewed the information in Section 2.2, "Neutron Fluence Methodology," of WCAP-14040, Revision 3 and determined that the proposed methodology adheres to the guidance of Regulatory Guide 1.190, and therefore, is acceptable.

The NRC staff has also reviewed WCAP-18124-NP-A Revision 0, "Fluence Determination with RAPTOR-M3G and FERRET," and determined that the proposed methodology adheres to the guidance of Regulatory Guide 1.190. The NRC therefore concluded that WCAP-18124-NP-A is acceptable for use in calculating reactor pressure vessel neutron fluence provided that the limitations and conditions listed in Section 4.0 of the NRC safety evaluation report were met.

Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," describes procedures acceptable to the NRC staff for calculating the effects of neutron irradiation embrittlement of the low-alloy steels of reactor pressure vessels.

In a February 27, 2004 safety evaluation, the NRC staff determined that the methodology described for determining material adjusted reference temperature values in WCAP-14040, Revision 3 was consistent with the guidance provided in the ASME Code, Standard Review Plan Section 5.3.1, "Reactor Vessel Materials," and Regulatory Guide 1.99, Revision 2, and was, therefore, acceptable.

Based on the foregoing, the proposed amendment will continue to ensure compliance with the above referenced regulations or guidance and will ensure that the functional capabilities or performance levels of equipment required for safe operation are met.

4.2 Significant Hazards Consideration The proposed amendment would revise Technical Specification 5.6.4, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," to update the specified analytical methods used to determine reactor coolant system pressure and temperature limits for operation of the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2). More specifically, the proposed amendment would allow use of the WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," for the development of reactor coolant system pressure and temperature limits and would allow WCAP-18124-NP-A, Revision 0, "Fluence Determination with RAPTOR-M3G and

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 10 of 11 FERRET," July 2018, to be used as an alternative to Section 2.2 of WCAP-1040-A, Revision 4.

Energy Harbor Nuclear Corp. has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

Implementation of the analytical methods as proposed would continue to provide assurance that appropriate reactor coolant system pressure and temperature limits are established to preserve the integrity of the reactor coolant system. The proposed amendment is based on NRG-approved methodologies.

Ensuring appropriate reactor coolant system pressure and temperature limits are established will not adversely affect a structure, system, or component of the plant, plant operations, design functions, or analysis that verifies the capability of a structure, system, or component to perform a design function. Since there are no adverse effects on systems, structures, or components, the likelihood of a malfunction is not increased and consequences of previously evaluated accidents in the Updated Final Safety Analysis Report are not changed.

Therefore, the proposed amendment does not significantly increase the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Implementation of the analytical methods as proposed would continue to provide assurance that appropriate reactor coolant system pressure and temperature limits are established to preserve the integrity of the reactor coolant system. The proposed amendment is based on NRG-approved methodologies.

The proposed amendment does not involve a physical alteration of the plant (no new or different type of equipment will be installed). The proposed amendment does not change the design of structures, systems, or components of the plant; or create new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases. The proposed amendment would continue to ensure reactor coolant system integrity.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 11 of 11 Implementation of the analytical methods would continue to provide assurance that appropriate reactor coolant system pressure and temperature limits are established in accordance with NRG-approved methodologies. This ensures that the plant is operated within design limits and the margin of safety in the plant safety analysis is maintained.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, Energy Harbor Nuclear Corp. concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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

5.0 ENVIRONMENTAL CONSIDERATION

S A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Attachment Technical Specification Page Markups (2 pages follow)

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT (COLR) (continued)

WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON,"

WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology,"

WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO'."

As described in reference documents listed above, when an initial assumed power level of 102% of RATED THERMAL POWER is specified in a previously approved method, 100.6% of RATED THERMAL POWER may be used when input for reactor thermal power measurement of feedwater flow is by the leading edge flow meter (LEFM).

Caldon, Inc. Engineering Report-SOP, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM TM System" Caldon, Inc. Engineering Report-160P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM TM System"

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, Overpressure Protection System (OPPS) enable temperature, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and LCO 3.4.12, "Overpressure Protection System (OPPS)"

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.

WCAP-18124-NP-A, Revision 0, "Fluence Determination with RAPTOR-M3G and FERRET," July 2012, may be used as an alternative to Section 2.2 of WCAP-14040-A, Beaver Valley Units 1 and 2 5.6 - 3 Amendments ~ I 94

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

NRG Letter, "Beaver Valley Power Station, Units 1 and 2 Acceptance of Methodology for Referencing Pressure and Temperature Limits Report (TAC Nos. MB3319 and MB3320)," dated October 8, 2002.

WCAP 14040 NP A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

The methodology listed in WCAP 14040 NP A was used with two exceptions:

  • ASME Code Case N 640, "Alternative Reference Fracture Toughness for Development of P T Limits for Section XI, Division 1."
  • ASME,Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 1996 version.
c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.6.5 Post Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.6 Steam Generator (SG) Tube Inspection Report 5.6.6.1 Unit 1 SG Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.1, "Unit 1 SG Program." The report shall include:

a. The scope of inspections performed on each SG,
b. Degradation mechanisms found,
c. Nondestructive examination techniques utilized for each degradation mechanism,
d. Location, orientation (if linear), and measured sizes (if available) of service-induced indications,
e. Number of tubes plugged during the inspection outage for each degradation mechanism,
f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator, and Beaver Valley Units 1 and 2 5.6 - 4 Amendments~ I 94

Enclosure B L-20-274 Justification of Using RAPTOR-M3G for Reactor Pressure Vessel Extended Beltline Materials at Beaver Valley Units 1 and 2 (16 pages follow)

@ Westinghouse Westinghouse Non-Proprietary Class 3 To: Benjamin Mays Date: June 18, 2020 Cc: Donald McNutt From: Nuclear Operations and Radiation Analysis Phone: (412) 374-2774 Our ref: LTR-REA-20-59-NP, Rev. 0 References 1. Nuclear Regulatory Commission (NRC) Regulatory Issue Summary (RIS) 2014-11 ,

"Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," October 14, 2014. (ADMAS Accession Number ML14149A165)

2. 10 CFR Appendix G to Part 50, "Fracture Toughness Requirements."
3. Westinghouse Report, WCAP-18124-NP-A, Revision 0, "Fluence Determination with RAPTOR-M3G and FERRET," July 2018.
4. Nuclear Regulatory Commission Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.
5. RSICC Data Library Collection DLC-185 , "BUGLE-96 Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996.
6. RSICC Computer Code Collection PSR-145 "FERRET: Least-Squares Solution to Nuclear Data and Reactor Physics Problems," January 1980.
7. RSICC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross Section Compendium," July 1994.
8. Pressurized Water Reactor Owners Group (PWROG) Report, PWROG-15109-NP-A, Revision 0, "PWR Pressure Vessel Nozzle Appendix G Evaluation," January 2020.

Subject:

Justification of Using RAPTOR-M3G for Reactor Pressure Vessel Extended Beltline Materials at Beaver Valley Units 1 and 2 1 Introduction RAPTOR-M3G was used in the fast neutron (E > 1.0 MeV) fluence evaluation of Capsule Y for Beaver Valley Unit 2. Per the requirement of Regulatory Issue Summary (RIS) 2014-11 (Reference 1), the reactor pressure vessel (RPV) nozzles and any RPV materials that exceed a fast neutron (E > 1. 0 Me V) fluence of IE+ 17 n/cm2 at the end-of-license extension (EOLE) must be evaluated with respect to fracture toughness.

Similarly, RAPTOR-M3G was used to evaluate fast neutron fluence for Beaver Valley Unit 1 RPV materials.

The RPV materials evaluated for Beaver Valley Units 1 and 2 are listed in Table 1. Certain materials in this table were determined to have fluence values less than IE+ 17 n/cm2 and therefore did not require specific evaluation with respect to neutron embrittlement and fracture toughness.

The active core height extends from -182.88 cm to +182.88 cm in the RAPTOR-M3G model and that represents the axial extent of the traditional beltline region of the reactor vessel which, by definition, is the

© 2020 Westinghouse Electric Company LLC All Rights Reserved

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Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 2 of 16 June 18, 2020 region that directly surrounds the effective height of the active core (Reference 2). When compared to the axial elevations of the RPV materials evaluated in Table 1, the intermediate shell, intermediate shell longitudinal welds, intermediate shell to lower shell circumferential weld, lower shell, and lower shell longitudinal welds are categorized as the traditional beltline region. Therefore, these materials have been approved by the Nuclear Regulatory Commission (NRC) for generic application of RAPTOR-M3G for fast neutron (E > 1.0 MeV) fluence determination per WCAP-18124-NP-A (Reference 3).

The 54 effective full-power years (EFPY) fast neutron (E > 1.0 MeV) fluence at the EOLE for the RPV materials in the extended beltline region for Beaver Valley Unit 1 and Unit 2 using RAPTOR-M3G are listed in Table 2. The limitations and conditions for using RAPTOR-M3G for fast neutron fluence determination are stipulated in Section 4.0 of the safety evaluation letter captured in Reference 3. The limitations and conditions of Reference 3:

1. Applicability of WCAP-18124-NP, Revision 0, is limited to the RPV region near the active height of the core based on the uncertainty analysis performed and the measurement data provided.

Additional justification should be provided via additional benchmarking, fluence sensitivity analysis to response parameters of interest (e.g., pressure-temperature limits, material stress/strain), margin assessment, or a combination thereof, for applications of the method to components including, but not limited to, the RPV upper circumferential weld and reactor coolant system inlet and outlet nozzles and reactor vessel internal components.

2. Least squares adjustment is acceptable if the adjustments to the MIC ratios and to the calculated spectra values are within the assigned uncertainties of the calculated spectra, the dosimetry measured reaction rates, and the dosimetry reaction cross sections. Should this not be the case, the user should re-examine both measured and calculated values for possible errors. If errors cannot be found, the particular values causing the inconsistency should be disqualified.

The second limitation and condition listed above does not apply as the least-squares procedures were not used to adjust the calculated fast neutron (E > 1.0 MeV) fluence values for RPV materials evaluated in the reactor vessel integrity analysis. The least-squares results were only used to compare the calculations and measurements from the evaluated dosimetry and validate the neutron transport models, and those comparisons showed satisfactory results.

The Reference 3 neutron fluence methodology, however, was used to determine the fast neutron (E > 1.0 MeV) fluence values for RPV materials in the extended beltline region. Therefore, additional justification should be provided for use of the methodology in this region. Per Reference 4, analytical uncertainty analysis and measurement benchmark are two main components required to qualify a neutron fluence calculational methodology for any applications outside of the approved scope of applicability. The following information provides the additional justification of using RAPTOR-M3G for fast neutron fluence determination for the RPV extended beltline regions at Beaver Valley Units 1 and 2 by summarizing a comprehensive analytical uncertainty analysis for reactor pressure vessel extended beltline materials and the additional benchmark data from a Westinghouse 4-loop pressurized water reactor (PWR) in the extended beltline region.

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Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 3 of 16 June 18, 2020 Table 1: Axial Locations for Reactor Vessel Materials Unit 1 Axial Location1 11 Unit 2 Axial Location1 11 Material [cm] [cm]

Outlet Nozzle to Nozzle Shell Welds 260.88 262.31 Lowest Extent Inlet Nozzle to Nozzle Shell Welds 255.17 251.44 Lowest Extent Nozzle Shell Lowest Extent 203.71 211.60 Nozzle Shell to Intermediate Shell 203.71 211.60 Circumferential Weld Intermediate Shell 121 -52.07 to 203.71 -49.27 to 211.60 Intermediate Shell Longitudinal Weldsl21 -52.07 to 203.71 -49.27 to 211.60 Intermediate Shell to Lower Shell -52.07 -49.27 Circumferential Weldl 1 2 Lower Shell 121 -307.59 to -52.07 -309.34 to -49.27 Lower Shell Longitudinal Weldsl 1 2

-307.59 to -52.07 -309.34 to -49.27 Lower Shell to Lower Vessel Head -307.59 -309.34 Circumferential Weld Notes:

1. Values listed are indexed to Z = 0.0 cm at the midplane of the active fuel stack.
2. RPV materials with fast neutron fluence values reported at axial locations between

-182.88 cm to+ 182.88 cm (traditional beltline materials).

Table 2: Fast Neutron (E > 1.0 MeV) Fluence at 54 EFPY for the Extended Beltline Region Materials Material Unit 1 [n/cm 2 ] Unit 2 [n/cm 2 ]

Outlet Nozzle to Nozzle Shell Welds l.05E+l7 5.96E+ 16 Lowest Extent Inlet Nozzle to Nozzle Shell Welds l.47E+ 17 l.39E+ 17 Lowest Extent Nozzle Shell Lowest Extent 7.11E+l8 4.21E+ 18 Nozzle Shell to Intermediate Shell 7.11E+l8 4.21E+ 18 Circumferential Weld Lower Shell to Lower Vessel Head l.30E+l6 l.89E+ 16 Circumferential Weld

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Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 4 of 16 June 18, 2020 2 Additional Benchmarking Measurements To collect measurement benchmark data for the extended beltline region, ex-vessel neutron dosimetry (EVND) was irradiated at the elevation of the reactor vessel supports at a Westinghouse 4-loop plant. The elevation of the reactor vessel supports is approximately 8.5 feet above the core midplane. The specific axial locations of the EVND capsules to the core midplane (Z = 0.0 cm) and time of irradiation are listed in Table 1. The dosimetry sensors included in these EVND capsules are listed in Table 2. The measured dosimetry reactions for those sensors are listed in Table 3.

Table 1: Location and Time of Irradiation for Sensor Sets Analyzed at RPV Supports Azimuthal Axial Elevation Cycle(s) of Capsule ID Sensor Location Location [cm] Irradiation E Ex-Vessel 180° 257.99 11 A Ex-Vessel 225° 255.75 11 K Ex-Vessel 180° 257.99 12-19 Table 2: Sensor Set Contents in EVND at RPV Supports Radiometric Monitor Capsule ID Fe Ni Cu Ti Co Nb U-238 Np-237 E X X X X X -- X X A X X X X X -- X X K X X X X X X -- --

Table 3: Measured Dosimetry Reactions in EVND at RPV Supports Material Reaction of Interest Neutron Energy ResponseC1> Product Half-LifeC2)

Copper 63 Cu (n,a) 6°Co 4.53-11.0 MeV 5.271 y Titanium 46Ti (n,p) 46Sc 3.70-9.43 MeV 83.788 d Iron 54pe (n,p) 54Mn 2.27-7.54 MeV 312.13 d Nickel 58Ni (n,p) 58Co 1.98-7.51 MeV 70.86 d 23su 238 U (n,f) m es 1.44-6.69 MeV 30.05 y Niobium 93Nb (n,n') 93mNb 0.95-5.79 MeV 16.13 y 237Np 237Np (n,f) m es 0.68-5.61 MeV 30.05 y Cobalt - Al 59Co (n,g) 6oco Thermal 5.271 y Note(s):

(1) Energies between which 90% of activity is produced (235 U fission spectrum) per ASTM E844-18.

(2) Half-life data is from ASTM El005-16.

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Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 5 of 16 June 18, 2020 2.1 Additional Benchmarking Neutron Transport Calculations The neutron transport calculations for the additional benchmarking at the 4-loop Westinghouse plant extended beltline region followed the Westinghouse fluence methodology described in Reference 3, which is the same methodology used for the neutron transport calculations performed in support of Beaver Valley Unit 1 RPV fluence evaluation and Beaver Valley Unit 2 Capsule Y. In the application of this methodology to the fast neutron exposure evaluations for the 4-loop Westinghouse plant EVND dosimetry sets at the RPV supports, forward transport calculations were carried out to directly solve for the space- and energy-dependent scalar flux, rp(r, 0,z,E).

For the additional benchmark analysis, the transport calculations were carried out using the RAPTOR-M3G three-dimensional discrete ordinates code and the BUGLE-96 (Reference 5) cross-section library. The BUGLE-96 library provides a 67-group coupled neutron-gamma ray group cross-section data set produced specifically for light water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P3 Legendre expansion and the angular discretization was modeled with an S20 order of angular quadrature.

A plan view of the reactor model is shown in Figure 1. In addition to the core, reactor internals, RPV, and concrete bioshield, the model also included explicit representations of the surveillance capsules, RPV clad, and RPV nozzles and supports. Section views of the reactor model are shown in Figure 2 and Figure 3.

In developing the model of the reactor geometry, nominal design dimensions were used for the various structural components. Water temperatures (and densities) in the core, bypass, and downcomer regions of the reactor were taken to be representative of full-power operating conditions. These coolant temperatures were varied on a cycle-specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, etc.

The r, 0,z geometric mesh description of the reactor model consisted of 241 radial by 190 azimuthal by 469 axial mesh intervals. Mesh sizes were chosen to ensure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion used in the calculations was 0.001.

The core power distributions used in the additional benchmarking transport analysis included fuel-assembly-specific initial enrichments, bumups, and axial power distributions. This information was used to develop spatial- and energy-dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of the fuel cycle-averaged neutron fluence rate, which, when multiplied by the appropriate fuel cycle length, provide the incremental fast neutron fluence exposure for each fuel cycle. The energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial 235 U enrichment and bumup history of the individual fuel assemblies. From the assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined. These fuel-assembly-specific neutron source strengths derived from the detailed isotopics were then converted from fuel pin Cartesian coordinates to the r,0,z spatial mesh arrays used in the RAPTOR-M3G discrete ordinates calculations.

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Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 6 of 16 June 18, 2020 Bioshield Bypass Figure 1: Reactor Geometry - Plan View at Core Midplane

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Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 7 of 16 June 18, 2020 Outlet Outlet Nozzle Plenum Down comer Upper Core Plate Core Modeled as Lower Core Plate Figure 2: Reactor Geometry - Section View at Outlet Nozzle Centerline

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Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 8 of 16 June 18, 2020 Inlet Nozzle Nozzle Support Figure 3: Reactor Geometry - Section View at Inlet Nozzle Centerline

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Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 9 of 16 June 18, 2020 2.2 Additional Benchmarking Dosimetry Evaluations The evaluations of the neutron sensor sets contained in the EVND dosimetry capsules at the 4-loop Westinghouse plant RPV supports followed the state-of-the-art least-squares dosimetry evaluation methodology described in Section 3.0 of Reference 3.

Least-squares adjustment methods provide the capability of combining the measurement data with the neutron transport calculation resulting in a best-estimate neutron energy spectrum with associated uncertainties. Best-estimates for key exposure parameters such as fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least-squares methods, as applied to reactor dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross sections, and the calculated neutron energy spectrum within their respective uncertainties.

For example, Ri +/- ()R; = L (a-ig +/- 8(JJ (¢g +/- 8¢)

g relates a set of measured reaction rates, Ri, to a single neutron spectrum, rpg, through the multigroup dosimeter reaction cross section, (Jig, each with an uncertainty b. The primary objective of the least-squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least-squares evaluation of the dosimetry, the FERRET code (Reference 6) was employed to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters along with associated uncertainties.

The application of the least-squares methodology requires the following input.

1. The calculated neutron energy spectrum and associated uncertainties at the measurement location.
2. The measured reaction rate and associated uncertainty for each sensor contained in the multiple foil set.
3. The energy-dependent dosimetry reaction cross sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the current application, the calculated neutron spectrum at each measurement location was obtained from the results of the previously-described additional benchmarking neutron transport calculations. The spectrum at each sensor set location was input in an absolute sense (rather than simply a relative spectral shape).

Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements. The sensor reaction rates were derived from the measured specific activities of each sensor set and the operating history of the respective fuel cycles. The dosimetry reaction cross sections were obtained from the SNLRML dosimetry cross section library (Reference 7).

In addition to the magnitude of the calculated neutron spectra, the measured sensor set reaction rates, and the dosimeter set reaction cross sections, the least-squares procedure requires uncertainty estimates for each of these input parameters. The following provides a summary of the uncertainties associated with the least-squares evaluation of the dosimetry.

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Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 10 of 16 June 18, 2020 2.3 Additional Benchmarking Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, the irradiation history corrections, and the corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM International consensus standards for reaction rate determinations for each sensor type.

After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input into the least-squares evaluation:

Reaction Uncertainty 63Cu (n,a) 6°Co 5%

46Ti (n,p) 46Sc 5%

54pe (n,p) 54Mn 5%

58Ni (n,p) 58Co 5%

238 U (n,f) m es 10%

93Nb (n,n')93mNb 10%

237Np (n,f) m es 10%

59Co (n,y) 6oco 35%*

These uncertainties are given at the 1a level.

2.4 Additional Benchmarking Dosimetry Cross Section Uncertainties As previously noted, the reaction rate cross sections used in the least-squares evaluations were taken from the SNLRML library. This data library provides reaction cross sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross sections and uncertainties are provided in a fine multi-group structure for use in least-squares adjustment applications. These cross sections were compiled from the ENDF/B-VI cross section evaluations and have been tested with respect to their accuracy and consistency for least-squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 Me V neutron sources.

Detailed discussions of the contents of the SNLRML library along with the evaluation process for each of the sensors are provided in Reference 7.

For sensors included in the dosimetry sets, the following uncertainties in the fission spectrum-averaged cross sections are provided in the SNLRML documentation package:

  • The cobalt content of older Co-Al foils used in EVND is not known for certain, but is believed to be between 0.438%

and 0.562%. To account for this unknown, the uncertainty assigned in the least-squares evaluations (typically 5%) was increased by roughly (0.562 / 0.438 = 1.28) to 5% + 28% = 33%. Rounded to the nearest five, an uncertainty of 35%

was input.

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Westinghouse Non-Proprietary Class 3 LTR-REA-2O-59-NP, Rev. 0 Page 11 of 16 June 18, 2020 Reaction Uncertainty 63eu (n,a) 60 eo 4.08-4.16%

46 Ti (n,p) 46 Sc 4.51-4.87%

54p e (n,p) 54Mn 3.05-3.11 %

58Ni (n,p) 58eo 4.49-4.56%

238 U (n,f) mes 0.54-0.64%

93Nb (n,n')93mNb 6.96-7.23%

237Np (n,f) mes 10.32-10.97%

s9eo (n,y) 6oeo 0.76-3.59%

These tabulated ranges provide an indication of the dosimetry cross section uncertainties associated with the sensor sets used in LWR irradiations.

2.5 Additional Benchmarking Calculated Neutron Spectrum Uncertainties While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks, and the dosimetry cross section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

Where Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg' and Rg specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

Where:

H = (g-g')2 2r2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range y (0 specifies the strength of the latter term). The value of 8 is 1.0 when g = g' and 0.0 otherwise.

The set of parameters defining the input covariance matrix for calculated spectra was as follows:

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Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 12 of 16 June 18, 2020 Flux Normalization Uncertainty (Rn) 15%

Flux Group Uncertainties (Rg, Rg,)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 25%

(E < 0.68 eV) 50%

Short-Range Correlation (0)

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range (y)

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 These uncertainty assignments are consistent with an industry consensus uncertainty of 15 - 20% (la) for the fast neutron portion of the spectrum and provide for a reasonable increase in the uncertainty for neutrons in the intermediate and thermal energy ranges.

2. 6 Additional Benchmarking Measurement-to-Calculation Comparison The comparison of the measurement results from each of the sensor set irradiations at RPV supports with corresponding analytical predictions at the measurement locations are presented in Table 4 and Table 5.

These comparisons are provided on two levels. On the first level, calculations of individual sensor reaction rates are compared directly with the measured data from the counting laboratories. This level of comparison is not impacted by the least-squares evaluations of the sensor sets. On the second level, calculated values of neutron exposure rates in terms of fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate are compared with the best-estimate exposure rates obtained from the least-squares evaluation.

In Table 4, comparisons of measurement-to-calculation (MIC) ratios are listed for the threshold sensors contained in the EVND dosimetry capsules irradiated at RPV supports that are approximately 8.5 feet above the core midplane. For the individual threshold foils, the average MIC ratio ranges from 0.62 to 1.28, with an overall average of 0.78 and an associated standard deviation of 25.5%. In this case, the overall average was based on an equal weighting of each of the sensor types with no adjustments made to account for the spectral coverage of the individual sensors.

In Table 5, best-estimate-to-calculation (BEIC) ratios for fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate resulting from the least-squares evaluation of the dosimetry sets is provided for the EVND capsules irradiated at the RPV supports, which are approximately 8.5 feet above the core midplane.

The BEIC ratio for the fast neutron (E > 1.0 MeV) fluence rate is 0.84 with an associated standard deviation of 8.9% and 0.93 with an associated standard deviation of 11 % for the iron atom displacement rate.

In summary, for the extended beltline region, the MIC data provided in Table 4 as well as the BEIC data provided in Table 5 suggest that the calculations are over predicting the neutron exposure, particularly at the high end of the energy spectrum. For instance, the bottom of the 90% neutron response for the copper,

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Westinghouse Non-Proprietary Class 3 LTR-REA-2O-59-NP, Rev. 0 Page 13 of 16 June 18, 2020 titanium, iron, and nickel dosimeters is 4.53 MeV, 3.70 MeV, 2.27 MeV, and 1.98 MeV, respectively.

Neutrons with energies greater than these constitute a small fraction of the neutron (E > 1.0 MeV) fluence rate in the extended beltline region. The BE/C values in Table 5 account for the spectral coverage of the different sensors, and provide an estimate of the key damage parameters, fast neutron (E > 1. 0 Me V) fluence rate and iron atom displacement rate, that result from an uncertainty-weighted reconciliation of all of the measurements and calculations. The BE/C values in Table 5 suggest that the calculated damage parameters are moderately conservative relative to the best-estimate values.

Table 4: Measured-to-Calculated (MIC) Reaction Rates -Ex-Vessel Capsule Located in the Vicinity of the RPV Supports Reaction Capsule E Capsule A Capsule K Average  % Std.Dev.

63 eu (n,a) 60 eo 0.65 - 0.58 0.62 8.0 46 Ti (n,p) 46 Sc 0.73 0.65 0.67 0.68 6.1 54pe (n,p) 54Mn 0.72 0.66 0.64 0.67 6.2 58Ni (n,p) 58 eo 0.75 0.67 0.65 0.69 7.7 238 U(ed) (n,f) mes 1.03 0.89 - 0.96 10.3 93Nb (n,n') 93mNb - - 1.28 1.28 -

237Np(ed) (n,f) mes 1.12 0.80 - 0.96 23.6 Average of MIC Results 0.78 25.5 Table 5: Best-Estimate-to-Calculated (BE/C) Exposure Rates - Ex-Vessel Capsule Located in the Vicinity of the RPV Supports Neutron (E > 1.0 MeV) Fluence Rate Iron Atom Displacement Rate Capsule BE/C BE/C E 0.91 0.99 A 0.76 0.82 K 0.84 1.00 Average 0.84 0.93

% Std.Dev. 8.9 11

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Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 14 of 16 June 18, 2020 3 Justification The uncertainty associated with a fluence determination methodology is comprised of two major components:

the results of an analytic uncertainty analysis and the results of benchmarking comparisons. An analytic uncertainty analysis assesses the level of confidence in key input parameters to a fluence calculation and quantifies the impact that plausible input parameter variations have on calculated fluence results.

Benchmarking comparisons refer to comparisons of fluence calculations performed with a candidate methodology to alternate calculations or to measurements from a representative environment.

A comprehensive analytical uncertainty analysis for the Beaver Valley Units 1 and 2 RPV extended beltline region demonstrates that the RAPTOR-M3G based calculations have a maximum of 42% uncertainty for the RPV extended beltline region of Beaver Valley Units 1 and 2. The estimated uncertainty for the nozzle shell and nozzle shell to intermediate shell circumferential weld is approximately 30%. Note that the uncertainty is increasing as the axial elevation of the RPV materials moving away from the active core. For elevations that are not as far from the top and bottom of the active core, the uncertainty of the fast neutron fluence determined using RAPTOR-M3G is much less than 42%. This completes an important part of qualifying RAPTOR-M3G as the fluence determination methodology for RPV extended beltline region per RG 1.190 (Reference 4) regulatory position 1. 4 .1.

Additional benchmarking work described in Section 2 was performed at a 4-loop Westinghouse plant near the RPV supports that are approximately 8.5 feet above the core midplane. This work concluded that the RAPTOR-M3G fluence determination methodology has about a 30% uncertainty in the fast neutron (E > 1.0 Me V) determination and the calculations typically overestimate the fast neutron (E > 1. 0 Me V) fluence and iron atom displacement (dpa) at the extended beltline region. This completes the second important part of qualifying RAPTOR-M3G as the fluence determination methodology for RPV extended beltline region per RG 1.190 (Reference 4) regulatory position 1.4.2.

The RPV extended beltline materials evaluated for Beaver Valley Units 1 and 2 in Table 2 are all located within an axial distance of 8.61 feet above or below the core midplane. However, the lower shell to lower vessel head circumferential weld is approximately 1.54 feet (47 cm) further away. It is important to note that this material is not classified as an extended beltline material since the projected fluence values for Beaver Valley Units 1 and 2 are well below IE+ 17 n/cm2 and therefore not evaluated with respect to fracture toughness. This circumferential weld has a calculated fast neutron (E > 1.0 MeV) fluence of 1.30E+ 16 n/cm2 for Beaver Valley Unit 1 and 1.89E+ 16 n/cm2 for Beaver Valley Unit 2 at 54 EFPY using RAPTOR-M3G, respectively, which both are more than a factor of 5 lower than the prescribed threshold of IE+ 17 n/cm2 for the definition of extended beltline region. Because the RAPTOR-M3G fluence determination methodology has a maximum of 42% uncertainty for the RPV extended beltline region this circumferential weld does not need to be included as extended beltline material that has to be evaluated for fracture toughness embrittlement effect.

From the discussion above, the methodology uncertainty for fast neutron fluence determination for these RPV extended beltline materials is also less than or equal to 42%. In addition, certain fast neutron (E > 1.0 MeV) fluence values used as input in the downstream fracture toughness evaluation have extra margin built-in. For example, in the past evaluations, the highest inlet nozzle to nozzle shell girth weld fast neutron (E > 1. 0 Me V) fluence value has been used for both inlet and outlet nozzles to nozzle shell girth welds, respectively, during the evaluation as this approach is conservative because the inlet nozzle to nozzle shell girth weld is closer to the core midplane. Finally, Section 3.4 of Reference 8 states that unless the fast neutron (E > 1.0 MeV) fluence for the nozzle material is greater than 4.28E+ 17 n/cm2 , embrittlement need not be considered for nozzle forging evaluation and the nozzles will be non-limiting compared to the beltline with respect to the

      • This record was final approved on 6/18/2020 11 :10:53 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 15 of 16 June 18, 2020 pressure-temperature limit curves. Embrittlement was conservatively considered in the reactor vessel integrity analyses for nozzle materials, even if the fluence was below this threshold. The fast neutron (E > 1.0 Me V) fluence values reported for both the inlet and outlet nozzles in Table 2 are more than a factor of 2 lower than this fast neutron (E > 1.0 MeV) fluence threshold. As the evaluated net RAPTOR-M3G methodology uncertainty is approximately 42% or less for this elevation based on additional benchmarking and analytical uncertainty analysis, it is also not credible that the inlet and outlet nozzle fast neutron (E > 1.0 MeV) fluence at the EOLE will exceed 4.28E+ 17 n/cm2 .

Additionally, previous reactor vessel integrity analysis showed that for Beaver Valley Unit 1 there is significant margin prior to these materials becoming limiting. For example, per WCAP-18102-NP, Appendix E, there is over 50 °F margin between the limiting extended beltline material RT PTs value and the limiting beltline material RTPTs value. Additionally, per WCAP-18102-NP, there is over 60 °F of margin between the limiting extended beltline material Adjusted Reference Temperature (ART) value and the limiting beltline material ART value. Therefore, based on the Beaver Valley Unit 1 reactor vessel integrity materials analysis, none of the extended beltline materials are limiting and these materials have significant margin before becoming the limiting materials.

For Beaver Valley Unit 2, the nozzle shell, nozzle shell longitudinal welds, and the nozzle shell to intermediate shell circumferential weld should be considered for facture toughness evaluation since some of these materials have initial RT NDT values and chemical compositions which could result in these materials becoming limiting compared to the beltline materials. The lowest extent of these materials is located at z =

211.66 cm, where the uncertainty of RAPTOR-M3G methodology is estimated to be approximately 30%.

Equation 6 in regulatory position 1.4.3 per RG 1.190 (Reference 4) is suggested to be used to apply to the calculated fast neutron fluence as the input to the downstream reactor integrity fracture toughness evaluation.

Equation 6 from RG 1.190 is listed below:

cf(%) - 20

</J = </Jc[l +Be+ 100 ]

Where Be is the bias factor of the methodology, which is zero, and ere is the calculational uncertainty of the methodology that is 30%. Therefore, a bias factor of 1.10 is suggested to be applied to the calculated fast neutron fluence for the nozzle shell, and nozzle shell and intermediate shell circumferential weld for the downstream reactor integrity fracture toughness evaluation.

Therefore, based on the analytical uncertainty analysis performed for extended beltline region using RAPTOR-M3G and additional benchmarking at the RPV extended beltline region summarized herein, and margin assessment provided in Reference 8 and the inherent margin between the extended beltline materials and beltline materials as analysed in the fluence valuation for Unit 1 and applying bias factor of 1.10 for Unit 2 limiting extended beltline region material, the RAPTOR-M3G fluence determination methodology is justified to be applicable to Beaver Valley Units 1 and 2 RPV extended beltline region fast neutron (E > 1.0 Me V) fluence determination for fracture toughness evaluation.

4 Application ofWCAP-18124-NP-A, Revision Oto the Extended Beltline Region - Conclusion Limitation and Condition #1 has been addressed in that the analytical uncertainty analysis and the additional benchmarking at the RPV extended beltline region summarized in Section 2, margin assessment documented in Reference 8, and the inherent margin between the extended beltline materials and beltline materials as analysed in the fluence valuation for Unit 1 and applying bias factor of 1.10 for Unit 2 limiting extended beltline region material, have provided additional justification supporting the use of the Reference 3

      • This record was final approved on 6/18/2020 11 :10:53 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LTR-REA-20-59-NP, Rev. 0 Page 16 of 16 June 18, 2020 methodology for the extended beltline regions of the Beaver Valley Units 1 and 2 RPV. Due to the inherent margin between the beltline and extended beltline for Beaver Valley Unit 1, no bias factor is necessary when implementing RAPTOR-M3G fluence. For Beaver Valley Unit 2, a bias factor of 1.10 is recommended to be used for fluence values determined using RAPTOR-M3G for the limiting extended beltline materials, nozzle shell and nozzle shell to intermediate shell circumferential weld.

If you have any questions, please contact the undersigned.

Electronically Approved* Electronically Approved*

Author: Jianwei Chen Reviewer: Eugene T. Hayes Nuclear Operations and Radiation Analysis Nuclear Operations and Radiation Analysis Electronically Approved*

Approver: Laurent P. Houssay, Manager Nuclear Operations and Radiation Analysis Electronically approved records are authenticated in the electronic document management system.

      • This record was final approved on 6/18/2020 11 :10:53 AM. (This statement was added by the PRIME system upon its validation)
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