LR-N23-0016, and Salem Generating Station, Units 1 and 2 - Report of Changes, Tests, and Experiments
| ML23059A046 | |
| Person / Time | |
|---|---|
| Site: | Salem, Hope Creek  |
| Issue date: | 02/28/2023 |
| From: | Jennings J Public Service Enterprise Group |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| LR-N23-0016 | |
| Download: ML23059A046 (1) | |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 0 PSEG NudearLLC 10 CFR 50.59(d)(2)
LR-N23-0016 February 28, 2023 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 Docket No. 50-354 Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DRP-70 and DRP-75 NRC Docket Nos. 50-272 and 50-311
Subject:
Report of Changes, Tests, and Experiments Per 10 CFR 50.59(d)(2), PSEG Nuclear LLC forwards summaries of changes, tests, and experiments implemented during the period of January 1, 2021 through December 31, 2022 at the Hope Creek and Salem Generating Stations There are no regulatory commitments contained in this letter.
Please contact Harry Balian at (856) 339 - 2173 if you have questions concerning this submittal.
Sincerely, Jennings, Digitally signed by Jennings, Jason Date: 2023.02.28 07:50:11 -05'00' Jason Jason Jennings Director, Regulatory Compliance Attachments
Document Control Desk Page 2 LR-N23-0016 cc: USNRC Regional Administrator Region 1 USNRC NRR Project Manager - Salem & Hope Creek USNRC Senior Resident Inspector - Hope Creek USNRC Senior Resident Inspector - Salem NJ Department of Environmental Protection, Bureau of Nuclear Engineering PSEG Nuclear Commitment Coordinator PSEG Nuclear Records Management
LR-N23-0016 Attachment 1 Hope Creek Generating Station Renewed Facility Operating License NPF-57 Docket No. 50-354 Report of Changes, Tests, and Experiments
LR-N23-0016 H2020-089, Hope Creek RACS Flooding SPV Mitigation FRC 2021-011 (80127117)
This activity is a permanent modification that allows operators to bypass the normal control circuit for reactor auxiliaries cooling system (RACS) heat exchanger inlet and outlet valves HV-2207 and HV-2346. The change also adds an alarm for RACS room flooded isolation signal and indication of activated bypass for the subject valves. The modification is necessary to eliminate single point vulnerabilities associated with loss of RACS cooling if a false isolation signal occurs. The isolation of RACS cooling could lead to a plant scram. The change allows operators to restore RACS cooling after confirming that the isolation signal is not valid; thereby avoiding an unnecessary scram.
An evaluation per 10 CFR 50.59 was required because adding the capability to bypass automatic increases the probability that HV-2207 or HV-2346 will malfunction. The evaluation determined the increased probability of malfunction is minimal. No other requirements of 10 CFR 50.59(c)(2)(i) are affected. FRC approved the evaluation in September 2021 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
H2021-004, Artificial Island Wind Port FRC F2021-006 and F2022-001 (80128565)
In April 2021, the PSEG Nuclear LLC (PSEG) fleet review committee (FRC) approved an evaluation of the proposed New Jersey (NJ) Wind Port facility (WO 80128565) that addressed both construction and operation of the NJ Wind Port.
PSEG revised the original 10 CFR 50.59 evaluation to delete evaluation of NJ Wind Port operation. The revised evaluation only considers the effects of construction activities. FRC approved the revised evaluation in January 2022 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
PSEG may issue new or further revise the existing 10 CFR 50.59 evaluation to review future phases of construction and inception of operations.
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LR-N23-0016 H2021-022, RACS Isolation SPV Mitigation FRC F2021-004 (80128187)
This activity is a permanent modification that allows operators to bypass the normal control circuit for reactor auxiliaries cooling system (RACS) containment isolation valves. The modification is necessary to eliminate single point vulnerabilities associated with loss of RACS cooling if a false isolation signal occurs. The isolation of RACS cooling could lead to a plant scram. The change includes primary containment isolation system (PCIS) indication of bypassed and inoperable containment isolation valves (CIV). Implementation of this change is partially complete.
An evaluation per 10 CFR 50.59 was required because adding the capability to bypass automatic CIV automatic closure adversely affects an updated final safety analysis report (UFSAR) design function. Specifically, the modification increases the probability that a system, structure, or component (SSC) will malfunction. The evaluation determined the increased probability of malfunction is minimal because the station enters a technical specification ACTION if the bypass is used. No other requirements of 10 CFR 50.59(c)(2)(i) are affected. FRC approved the evaluation in March 2021 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
H2021-071, TCCP 4HT 21-023 RPV/RWCU BHDL Isolation FRC F2022-004 and F2022-010 (80129316)
This activity is a temporary modification to isolate flow through the reactor water cleanup (RWCU) bottom head drain piping. Flow through the bottom head drain piping caused pipe wall thinning. Limiting flow through this line reduces flow accelerated corrosion (FAC) of the piping until a permanent repair is implemented. The change is not yet implemented.
An evaluation was required because the activity involves a change that adversely affects an updated final safety analysis report (UFSAR) described design function and a change to a procedure that adversely affects how UFSAR described design functions are performed. The evaluation determined that the activity satisfies the requirements of 10 CFR 50.59(c)(2)(i) through (viii). FRC approved the original evaluation in May 2022 and a revised evaluation in October 2022 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
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LR-N23-0016 H2021-094, Replacement of 20 kVA 1E Inverters 1CD481 & 1CD482 FRC F2021-013 (80128252)
This activity is a permanent modification to replace electric power inverters. The inverters, designated 1CD481 and 1CD482, provide regulated and uninterruptible electric power to class 1E instrument distribution panels. The original inverters were obsolete analog machines. The replacement inverters are digital machines verified to perform equal to or better than the original equipment.
The activity involves a change to an SSC the failure of which, when installed on multiple channels, could adversely affect an UFSAR described design function.
Because the new inverters are digitally controlled, when installed across multiple vital buses, there is a potential that a software common cause failure (SCCF) may make multiple channels unavailable at the same time. Although the new digital inverter is only installed in one safety related channel (Channel C), the screen was performed reflecting the intent to install the new inverters across all redundant safety related channels. For this reason, the 10 CFR 50.59 screen determined that a 10 CFR 50.59 Evaluation is required. The 10 CFR 50.59 evaluation relied upon a qualitative assessment of the new inverters performed per RIS 2002-22 Supplement 1- Clarification on Endorsement of NEI Guidance in Designing Digital Upgrades in Instrumentation and Control Systems and concluded that the new inverters do not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety.
The evaluation determined that the activity satisfies the requirements of 10 CFR 50.59(c)(2)(i) through (viii). FRC approved the evaluation in September 2021 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
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LR-N23-0016 Attachment 2 Salem Generating Station Renewed Facility Operating License DRP-70 and DRP-75 Docket Nos. 50-272 and 50-311 Report of Changes, Tests, and Experiments
LR-N23-0016 S2017-208 / S2017-158, Salem Chiller Replacement FRC F2018-06 (80111049 & 80123152)
This activity is a permanent modification to replace six chillers (three per unit). The Unit 1 chillers have been replaced and Unit 2 chiller replacements will begin in 2023. The chillers cool water that is used to support control area ventilation, emergency control air compressors, and other loads that are not safety related. The replacement chillers will function equal to or better than the original equipment. However, the replacement chillers use a digital control system requiring a safety evaluation.
The 10 CFR 50.59 evaluation relied upon a qualitative assessment performed per RIS 2002-22 Supplement 1- Clarification on Endorsement of NEI Guidance in Designing Digital Upgrades in Instrumentation and Control Systems and concluded that the replacement chillers do not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. FRC approved both evaluations in December 2018 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
S2021-003, Artificial Island Wind Port FRC F2021-006 and F2022-001 In April 2021, the PSEG Nuclear LLC (PSEG) fleet review committee (FRC) approved an evaluation of the proposed New Jersey (NJ) Wind Port facility (WO 80128565) that addressed both construction and operation of the NJ Wind Port.
PSEG revised the original 10 CFR 50.59 evaluation to delete evaluation of NJ Wind Port operation. The revised evaluation only considers the effects of construction activities. FRC approved the revised evaluation in January 2022 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
PSEG may issue new or further revise the existing 10 CFR 50.59 evaluation to review future phases of construction and inception of operations.
LR-N23-0016 S2021-024, 62K Lead Rod Burnup Limit FRC F2021-014 (WO 80106143)
This activity raised the allowable lead fuel burnup limit to 62,000 MWD/MTU from 60,000 MWD/MTU. This was an adverse change to an element of a UFSAR described evaluation methodology requiring evaluation under 10 CFR 50.59. NRC approved the methodology by letter to Westinghouse Electric Company in May 2006 (ADAMS Accession No. ML061420458). A license amendment was not required because Salem conforms to conditions in the NRC approval. FRC approved the evaluation in November 2021 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
S2021-101, Salem Unit 2 Upflow Conversion and Unit 1 and 2 Extended Leak Before Break (eLBB) Implementation - FRC F2021-12 (WO 80126494)
This is a permanent modification to the Unit 2 reactor internals that reversed the reactor coolant flow pattern in the baffle-barrel region from a down flow configuration to an up flow configuration. The Unit 2 reactor internals modification represented an adverse change to the UFSAR described design function of the reactor vessel internals to limit core bypass flow. This change required safety analyses to be re-run and other significant evaluations performed to demonstrate that the change is acceptable. These aspects of the proposed activity required a 10 CFR 50.59 evaluation. The evaluation was only required for the Unit 2 reactor internals modification because the extended leak before break (eLBB) was approved by license amendment. The 50.59 evaluation of the Unit 2 reactor internals modification determined that the reduction in core flow has minor impact on reactor coolant system (RCS) temperatures and overall pressure drop through the reactor vessel. Neither of these changes impact normal plant operation.
However, several safety analyses as described in the UFSAR are impacted. These include LOCA analyses, non-LOCA analyses, transients, LOCA mass and energy releases including containment response, and vessel and internals structural and thermal-hydraulic evaluations. Calculation revisions, evaluations, and assessments incorporated under the modification demonstrated that there is no adverse structural impact to the reactor vessel and internal components and that the event specific acceptance criteria for the affected safety analyses can be satisfied with baffle-barrel flow in the up flow configuration. FRC approved the evaluation in August 2021 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
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LR-N23-0016 S2021-102, Procedure Revision - S1(2).OP-SO.SW-0005 FRC F2022-03 (WO 80130347)
This activity is a change to standard operating procedures for the service water (SW) system that affect turbine driven auxiliary feedwater (TDAFW) pump ventilation. The TDAFW room cooler is an auxiliary building ventilation (ABV) component that is cooled by SW. The change revised S1(2).OP-SO.SW-0005, Service Water System Operation to remove the requirement to declare the TDAFW pump inoperable when TDAFW room cooler is unavailable. Having the room cooler (1(2)VHE36) out of service, but continuing to maintain the TDAFW pump operable adversely affects a UFSAR described design function resulting in a review under 10CFR50.59(c)(2).
The evaluation demonstrated that two AFW pumps remain operable through the duration of a design basis accident with the TDAFW room cooler unavailable under all design considerations and constraints. FRC approved the evaluation in March 2022 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
Subsequently, the NRC issued non-cited violation 05000272,05000311/2022004-01 because the evaluation was incorrect. The change resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component. PSEG restored the requirement to declare the TDAFW pump inoperable when TDAFW room cooler is unavailable and all affected documents are reverting to reflect the original prior to the activity.
S2021-202, Salem 2C Vital Inverter Replacement and 2 RMS Inverter Abandonment - FRC F2022-02 (80124612)
This activity is the replacement of the 2C Vital inverter and the abandonment of the 2C RMS inverter. Electrical loads powered by the 2C RMS inverter will be powered from 2C Vital Instrument Bus. The original inverters were obsolete analog machines. The replacement inverters are digital machines verified to perform equal to or better than the original equipment. Although the digital inverter is only being installed in one safety related channel (Channel c), the screen was performed assuming that the redundant safety related channels will eventually be replaced with the same digital inverters.
Because the new inverters are digitally controlled, when installed across multiple vital buses, there is a potential that a software common cause failure (SCCF) may make multiple channels unavailable at the same time. For this reason, the 10 CFR 50.59 Page 2
LR-N23-0016 screen determined that a 10 CFR 50.59 Evaluation is required. The 10 CFR 50.59 evaluation relied upon a qualitative assessment of the new inverters performed per RIS 2002-22 Supplement 1- Clarification on Endorsement of NEI Guidance in Designing Digital Upgrades in Instrumentation and Control Systems and concluded that the new inverters do not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety.
The 10 CFR 50.59 evaluation concluded that the proposed activity does not result in more than a minimal increase in the frequency of occurrence of an accident, does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety, does not result in more than a minimal increase in the consequences of an accident, does not result in more than a minimal increase in the consequences of a malfunction, does not create the possibility of an accident of a different type, does not create the possibility for a malfunction of an SC important to safety with a different result, and does not result in a design basis limit for a fission product barrier being exceeded or altered. FRC approved the evaluation in March 2022 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
S2022-072, 1(2)RH21 Leakby Evaluation (RWST Back-Leakage)
FRC 2022-05 (80131774)
This is an increase to the assumed back leakage through residual heat removal (RHR) isolation valve 1(2)RH21 during the recirculation phase of a loss of coolant accident.
The change affects calculated dose to the Salem control room and technical support center. Calculated dose at the exclusion area boundary and low population zone remain bounded by the UFSAR values. However, the evaluation determined that the change did not cause more than a minimal increase to the consequences of an accident and are based on existing methodology. FRC approved the evaluation in October 2022 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
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LR-N23-0016 S2022-088, Salem Unit 1 Upflow Conversion FRC F2022-07 (80130392)
This is a permanent modification to the Unit 1 reactor internals that will reverse the reactor coolant flow pattern in the baffle-barrel region from a down flow configuration to an up flow configuration.
The Unit 1 reactor internals modification represents an adverse change to the UFSAR described design function of the reactor vessel internals to limit core bypass flow. This change required safety analyses to be re-run and other significant evaluations performed to demonstrate that the change is acceptable. These aspects of the proposed activity required a 10 CFR 50.59 evaluation. The evaluation was only required for the Unit 1 reactor internals modification because the extended leak before break (eLBB) was approved by license amendment. The 50.59 evaluation of the Unit 1 reactor internals modification determined that the reduction in core flow has minor impact on reactor coolant system (RCS) temperatures and overall pressure drop through the reactor vessel. Neither of these changes impact normal plant operation.
However, several safety analyses as described in the UFSAR are impacted. These include LOCA analyses, non-LOCA analyses, transients, LOCA mass and energy releases including containment response, and vessel and internals structural and thermal-hydraulic evaluations. Calculation revisions, evaluations, and assessments incorporated under the modification demonstrated that there is no adverse structural impact to the reactor vessel and internal components and that the event specific acceptance criteria for the affected safety analyses can be satisfied with baffle-barrel flow in the up flow configuration. FRC approved the revised evaluation in July 2022 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii). The physical plant change is scheduled to occur in the Fall of 2023.
S2022-110, Procedure Revisions FRC F2022-09 (70224551)
This activity is a change to the operating procedures for the service water (SW) system that affects SW pump operability if the associated SW traveling water screen (TWS) is not functioning. The change revised S1(2).OP-SO.SW-0005, Service Water System Operation to replace the term inoperable with non-functioning when referring to traveling water screens. SW pump operability is not dependent on rotation of the associated TWS.
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LR-N23-0016 The evaluation demonstrated that the TWS provide a passive function to maintain structural integrity. FRC approved the evaluation in August 2022 after finding the proposed activities satisfied the requirements of 10 CFR 50.59(c)(2)(i) through (viii).
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