ML22300A084

Pressure and Temperature Limits Report (Ptlr), Revision 8

ML22300A084
Person / Time
Site:	Braidwood
Issue date:	11/10/2022
From:	Joel Wiebe NRC/NRR/DORL/LPL3
To:
References
Download: ML22300A084 (1)

Text

November 10, 2022 MEMORANDUM TO: File FROM: Joel S. Wiebe, Senior Project Manager /RA/

Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

SUBJECT:

BRAIDWOOD STATION, UNIT 2 - PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR), REVISION 8 By letter dated October 27, 2021 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML21300A075), Exelon Generation Company, LLC.

submitted the Braidwood, Unit 2 - pressure and temperature limits report (PTLR), Revision 8.

On February 1, 2022 (ML22032A333), Exelon Generation Company, LLC was renamed Constellation Energy Generation, LLC (Constellation, the licensee). The PTLR was submitted in accordance with Technical Specification (TS) 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR).

Braidwood Station, Unit 2, TS 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), was revised to allow the use of AREVA NP Topical Report BAW-2308, Revisions 1-A and 2-A, Initial RTNDT of Linde 80 Weld Materials for determining RCS pressure-temperature limits.

The licensee is not required to request approval of the PTLR and did not request approval.

However, the NRC staff performed an assessment of the PTLR. Based on the information submitted, the NRC staff finds that the pressure-temperature limits and low temperature over pressure setpoint limits in the PTLR, Revision 8, are acceptable.

Docket No. 50-457

Enclosures:

1. Email from Technical Staff

2. Assessment CONTACT: Joel S. Wiebe, NRR/DORL 301-415-6606

Email Dated October 2, 2022 Enclosure 1

From: Joel Wiebe To: Joel Wiebe

Subject:

FW: Braidwood Unit 2-- Staff Assessment of Pressure Temperature Limits Report (PTLR) Revision 8 Date: Tuesday, October 25, 2022 12:59:15 PM Attachments: Staff Assessment 9-5-2022.docx From: John Tsao <John.Tsao@nrc.gov>

Sent: Sunday, October 02, 2022 4:32 PM To: Joel Wiebe <Joel.Wiebe@nrc.gov>

Cc: Angie Buford <Angela.Buford@nrc.gov>

Subject:

Braidwood Unit 2-- Staff Assessment of Pressure Temperature Limits Report (PTLR) Revision 8

Joel, By letter dated October 27, 2021 (Agencywide Document Access and Management System Accession Number ML21300A075), Exelon Generation Company, LLC (the licensee) submitted to the United States Nuclear Regulatory Commission (NRC) Revision 8 of the pressure temperature limits report (PTLR) for Braidwood Station, Unit 2 in accordance with Technical Specification (TS) 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)." The licensee submitted the PTLR for information only. Exelon Generation Company, LLC, has been changed to Constellation Energy Generation, LLC.

Based on the information submitted, the NRC staff finds that the P/T limits and LTOP setpoint limits in the PTLR, Revision 8, acceptable.

As stated above, the licensee submitted the PTLR for information only. The attachment contains staffs assessment, not a safety evaluation.

As you and I had discussed prviously, you will disposition the attached assessment to file. Based on the instruction from my supervisor, Angie Buford, I am forwarding you the attached assessment directly. The attached assessment completes the effort by the Vessels and Internals Branch, Division of New and Renewed Licenses.

Thanks John Received: from BLAPR09MB6340.namprd09.prod.outlook.com (2603:10b6:208:2a3::19) by MN2PR09MB4971.namprd09.prod.outlook.com with HTTPS;

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Subject:

Braidwood Unit 2-- Staff Assessment of Pressure Temperature Limits Report (PTLR) Revision 8 Thread-Topic: Braidwood Unit 2-- Staff Assessment of Pressure Temperature Limits Report (PTLR) Revision 8 Thread-Index: AdjWnRv+mVvlGmgET3mBc1YmocGk6w==

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STAFF ASSESSMENT PRESSURE TEMPERATURE LIMITS REPORT, REVISION 8 TECHNICAL SPECIFICATIONS 5.6.6 BRAIDWOOD STATION UNIT 2 CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-457

1.0 INTRODUCTION

By letter dated October 27, 2021 (Agencywide Document Access and Management System (ADAMS) Accession No. ML21300A075), Exelon Generation Company, LLC) submitted to the U. S. Nuclear Regulatory Commission (NRC) Revision 8 of the pressure temperature limits report (PTLR) for Braidwood Station, Unit 2, in accordance with Technical Specification (TS) 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR).

On February 1, 2022 (ML22032A333), Exelon Generation Company, LLC was renamed Constellation Energy Generation, LLC (Constellation, the licensee). The licensee submitted the PTLR for information only.

2.0 REGULATORY EVALUATION

The NRC has established requirements in Title 10 of the Code of Federal Regulations (10 CFR) part 50) to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The NRC staff evaluates the acceptability of a facilitys proposed pressure and temperature (P/T) limits based on the NRC regulations and guidance discussed in the following subsections.

2.1 Regulatory Requirements The regulations in 10 CFR 50.36, Technical specifications, paragraph (a), requires that each operating license application for a production or utilization facility include proposed TSs and a summary statement of the bases for such specifications. Paragraph (c) of 10 CFR 50.36 requires, in part, that TSs include the following categories related to facility operation: (1) safety limits, limiting safety systems settings, and control settings; (2) limiting conditions for operations (LCO); (3) surveillance requirements; (4) design features; and (5) administrative controls.

Section 50.60 of 10 CFR Acceptance criteria for fracture prevention measures for light water nuclear power reactors for normal operation imposes fracture toughness and material embrittlement surveillance program requirements set forth in appendices G and H to 10 CFR part 50.

Appendix G, Fracture Toughness Requirements, to 10 CFR part 50 requires, in part, that facility P/T limits for the reactor pressure vessel (RPV) be at least as conservative as those obtained by applying the linear elastic fracture mechanics methodology of appendix G, Fracture Toughness Criteria for Protection Against Failure, to section XI of the American Society for Mechanical Engineers Boiler and Pressure Vessel Code.

Enclosure 2

Appendix H, Reactor Vessel Material Surveillance Program Requirements, to 10 CFR part 50 establishes requirements for a facilitys surveillance program for monitoring fracture toughness due to neutron irradiation.

Generic Letter (GL) 96-03, Relocation of Pressure and Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, dated January 31, 1996 (ML03111004) permits relocation of the P/T limits from the TS to a PTLR. GL 96-03 recommends that licensees who seek a license amendment for relocation (1) generate their P/T limits in accordance with an NRC-approved methodology, (2) comply with 10 CFR part 50, appendices G and H, (3) reference NRC-approved methodologies in the TS, (4) define the PTLR in TSs section 1.0, (5) develop a PTLR to contain the P/T limit curves, and (6) modify applicable sections of the TS accordingly.

2.2 Regulatory Guidance Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, dated May 1988 (Reference 2), contains guidance for RPV embrittlement integrity evaluations.

RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, dated March 2001 (Reference 3), describes methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence with respect to the general design criteria (GDC) in appendix A to 10 CFR part 50. In consideration of the guidance set forth in RG 1.190, GDCs 14, 30, and 31, are applicable (see section 2.1 of this safety evaluation (SE)).

Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, dated October 14, 2014, clarifies that P/T limits for ferritic RPV components, such as RPV inlet and outlet nozzles, could be more limiting because higher stress levels from structural discontinuities could result in a lower allowable pressure. RIS 2014-11 also clarifies that the RPV beltline definition in appendix G to 10 CFR part 50 is applicable to all RPV ferritic materials with projected neutron fluence values greater than 1 x 1017 neutrons per square centimeters (n/cm2)

(E > 1.0 MeV), and that this fluence threshold remains applicable for the design life as well as throughout the licensed operating period of the reactor.

By letter dated March 21, 2002 (ML020800488), the NRC approved the use of Technical Specifications Task Force (TSTF) Traveler TSTF-419-A, Revision 0, Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR (ML012690234). By letter dated August 4, 2011 (ML110660285), the NRC staff clarified the use of TSTF-419-A.

Branch Technical Position (BTP) 5-2, Revision 3 (Reference 5), Overpressurization Protection of Pressurized-Water Reactors while Operating at Low Temperatures, in NUREG-0800, provide guidance to the NRC staff in reviewing overpressurization protection of pressurized-water reactors (PWRs) while operating at low temperatures. Paragraph B.1 of BTP 5-2 specifies that the low temperature overpressure protection (LTOP) system be capable of relieving pressure during all anticipated overpressurization events at a rate sufficient to satisfy the TS limits while operating at low temperatures.

3.0 Staff Assessment The NRC staff reviewed Braidwood, Unit 2, PTLR, Revision 8 (ML ML21300A076), which includes P/T limit curves, LTOP system curves, and reactor vessel material surveillance program.

GL 96-03 stipulates that licensees can submit to the NRC an updated PTLR for information only without requesting for a license amendment when the NRC has approved the original PTLR for implementation in the plant TS and that the updated P/T limit curves are developed by the NRC-approved methodology as referenced in the TS.

Braidwood, Unit 2, TS 5.6.6 requires that:

TS 5.6.6.a. RCS P/T limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates, and power operated relief valve (PORV) lift settings 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, Low Temperature Overpressure Protection (LTOP) System, TS 5.6.6.b. The analytical methods used to determine the RCS P/T limits shall be those previously reviewed and approved by the NRC, specifically, those described in the following documents:

1. NRC letters dated January 21, 1998, Byron Station Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report,

2. NRC letter dated August 8, 2001, Issuance of Exemption from the requirements of 10 CFR 50.60 and Appendix G for Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2,

3. Westinghouse WCAP-16143, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2,

4. NRC letter dated August 31, 2020 [ML20022A336], Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2, Exemption from the Requirements of 10 CFR 50.61 and 10 CFR 50, Appendix G (EPID L-2019-LLE-0022), and NRC letter dated September 18, 2020 [ML20163A046], Braidwood Station, Units 1 and 2, and Byron Station Unit Nos. 1 and 2 -Issuance of Amendment Nos. 217, 217, 221, and 221 Regarding Reactor Coolant System Pressure and Temperature Limits Report Technical Specifications (EPID L-2019-LLA-0215), and

5. The PTLR will contain the complete identification for each of the TS referenced Topical Reports used to prepare the PTLR (i.e., report no., title, revision, date, and any supplements); and TS 5.6.6.c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Braidwood, Unit 2, PTLR, Revision 8 (Updated PTLR)

As stated in section 2 of the PTLR, Revision 8, the licensee developed the updated PTLR for Braidwood, Unit 2, using the methodology specified in TS 5.6.6.

The licensee stated that WCAP-18370-NP, Revision 0, Braidwood Units 1 and 2 Heatup and Cooldown Limits for Normal Operation, June 2019, provides the basis for the Braidwood, Unit 2, P/T curves for the 57 effective full-power year (EFPY), along with the best estimate chemical compositions, fluence projections, and adjusted reference temperatures, used to determine these limits. The licensee developed the P/T curves as shown in WCAP-18370-NP, Revision 0, based on the methods in WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et al., May 2004, with the following two exceptions:

a) Elimination of the [reactor vessel] flange requirements documented in WCAP-16143-P, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2, W. Bamford, et al., October 2014.

b) The initial reference temperatures of the inlet/outlet nozzle forging to shell welds are determined using BAW-2308, Initial RTNDT of Linde 80 Weld Materials, August 2005 and March 2008, in lieu of the ASME NB-2300 requirements.

The WCAP-16143-P report documents the technical basis for the elimination of the reactor vessel flange requirements. The NRC approved the use of WCAP-16143-P by letter dated October 28, 2015 (ML15232A441), from J. S. Wiebe, NRR, to B.C. Hanson, Exelon Generation Company, LLC, Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2 - Issuance of Amendments to Utilize WCAP-16143-P, Revision 1, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2.

The licensee used the Master Curve fracture toughness properties from BAW-2308, Revision 1-A and Revision 2-A for the inlet/outlet nozzle to upper shell forgings welds in lieu of the methods in WCAP-14040-A.

By letter dated September 18, 2020 (ML20163A046), NRC approved the use of BAW-2308 and updated TS 5.6.6 as shown in NRC letter from J. S. Wiebe, NRR, to B.C. Hanson, Exelon Generation Company, LLC, Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2- Issuance of Amendment Nos. 217, 217, 221, and 221 Regarding Reactor Coolant System Pressure and Temperature Limits Report Technical Specifications (EPID L-20 l 9-LLA-0215). Subsequently, the licensee revised Braidwood Station, Unit 2, TS 5.6.6 to allow the use of BAW-2308, revisions 1-A and 2-A, for determining RCS pressure-temperature limits.

In addition to the NRCs letters dated August 31, 2020, and September 18, 2020, by letter dated November 22, 2006 (ML061890003), the NRC approved the use of BAW-2308.

The NRC staff has approved the use of the two exceptions in developing the P/T curves as shown in the above NRC letters. Therefore, the licensees two exceptions from WCAP-14040-A are acceptable. In addition, by the NRCs letter dated September 18, 2020, the NRC approved the licensees update of TS 5.6.6 which includes documents associated with the two exceptions.

TS 5.6.6.b.3 includes the use of WCAP-16143 and TS 5.6.6.b.4 includes the used of BAW-2308.

The NRC staff performed an independent calculation and verified the P/T curves in figure 2.1 of the PTLR, Revision 8, which contains the heatup curve, the leak test curve, and criticality curve.

The NRC staff also verified the cooldown curves in figure 2.2 of the PTLR, Revision 8.

The NRC staff determined that the P/T curves for the 57 EFPYs as presented in figures 2.1 and 2.2 of the updated PTLR are developed based on the methodology that are referenced in TS 5.6.6. Therefore, the NRC staff finds that the updated P/T curves in figures 2.1 and 2.2 are acceptable.

LTOP System The nominal lift settings for PORVs are shown in figure 3.1 and table 3.1 of the Braidwood, Unit 2, PTLR, Revision 8. The licensee stated that these limits are based on TR-SCS-19-14, Revision 1, Braidwood Units I and 2 Low Temperature Overpressure Protection System (LTOPs) Analysis for 57 EFPY, September 19, 2019. The LTOP setpoints are based on P/T limits that were established in accordance with 10 CFR 50, appendix G, without allowance for instrumentation error. The topical report WCAP-14040-A, Revision 4, contains methodology for the generation of the LTOP setpoints. The licensee stated that the LTOP PORV nominal lift settings shown in figure 3.1 and table 3.1 of the updated PTLR account for appropriate instrument error.

The licensee stated that Braidwood, Unit 2, procedures governing the heatup and cooldown of the RCS require the arming of the LTOP system for RCS temperature less than 350°F and disarming of LTOP for RCS temperature of 350 degree Fahrenheit (°F) and above. The licensee stated that the last LTOP PORV segment in table 3.1 of the PTLR extends to 400 °F where the pressure setpoint is 2335 pounds per square inch guage. The licensee stated that this is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.

The NRC staff notes that the purpose of the LTOP is to minimize the reactor vessel shell and other RCS components (e.g., reactor coolant pump No. 1 seal and RCS piping) being overstressed by excessive pressure during low temperature operation. As such, when RCS temperature is less than 350 °F and the RCS pressure exceeds the PORV setpoints, the PORV will lift to release the pressure; thereby reduces stresses on the reactor vessel shell.

WCAP-14040-A describes the derivation of the LTOP setpoints. The NRC staff finds that the LTOP setpoints were derived based on the NRC-approved topical report WCAP-14040-A and, therefore, are acceptable.

Reactor Vessel Boltup Temperature (Non-Technical Specification)

The licensee stated that the minimum boltup temperature for the reactor vessel flange shall be 60 °F. The licensee stated that boltup is a condition in which the reactor vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere. The NRC staff finds the minimum boltup temperature for the reactor vessel flange 60 °F is consistent with the P/T curves and 10 CFR 50, appendix G, and is, therefore, acceptable.

Reactor Vessel Material Surveillance Program The licensee stated that its pressure vessel material surveillance program is in compliance with appendix H to 10 CFR 50, Reactor Vessel Radiation Surveillance Program. The material test requirements and the acceptance standards use the reference nil-ductility temperature, RTNDT, which is determined in accordance with American Society of Mechanical Engineers (ASME)

Code, section III, NB-2331.

The NRC staff notes that the empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with the ASME Code, section XI, appendix G, Protection Against Non-Ductile Failure. The licensee stated that the surveillance capsule removal schedule meets the requirements of ASTM E185-82. The licensee further stated that the fourth reactor vessel material irradiation surveillance specimens (Capsule V) have been analyzed to determine changes in material properties as documented in WCAP-18107-NP, Revision 0, Analysis of Capsule V from the Exelon Generation Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program, May 2016. The licensee has completed its surveillance capsule testing for the original 40-year operating period. The licensee removed the remaining two capsules, Y and Z, and placed them in the spent fuel pool to avoid excessive fluence accumulation should they be needed to support life extension. The removal summary is provided in table 4.1 of the PTLR.

The NRC staff determined that the licensees reactor vessel material surveillance program is acceptable because it is in compliance with the 10 CFR 50, appendix H. The NRC staff further determined that the licensee has appropriately withdrawn the surveillance capsules in accordance with ASTM E185-82 pursuant to 10 CFR 50, appendix H.

4.0 CONCLUSION

Based on the above assessment, the NRC staff finds that the P/T limits and LTOP setpoint limits in the PTLR, Revision 8, acceptable.

Principal Contractor: JTsao NRR