ML22063A450
| ML22063A450 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/04/2022 |
| From: | Shawn Williams Plant Licensing Branch II |
| To: | Grzeck L Duke Energy Carolinas |
| Williams S | |
| References | |
| L-2021-LLA-0157 | |
| Download: ML22063A450 (9) | |
Text
From: Williams, Shawn To: Grzeck, Lee Cc: Vaughan, Jordan L
Subject:
Oconee Nuclear Station, Units 1, 2, and 3 - Request for Additional Information RE: Application to Revise TS 3.7.7, Low Pressure Service Water (LPSW) System, to Extend the Completion Time for One Required Inoperable LPSW Pump (EPID L-2021-LLA-0157)
Date: Friday, March 04, 2022 11:19:26 AM Attachments: Oconee TS 3.7.7 RAIs dated March 4, 2022.docx
Dear Mr. Grzeck,
By [[letter::RA-21-0005, Application to Revise Technical Specification 3.7.7, Low Pressure Service Water (LPSW) System, to Extend the Completion Time for One Required Inoperable LPSW Pump on a Temporary Basis|letter dated September 2, 2021]], Duke Energy Carolinas, LLC, submitted a license amendment request for the Oconee Nuclear Station, Units 1, 2, and 3 to revise Technical Specification (TS) 3.7.7, Low Pressure Service Water (LPSW) System, to extend the Completion Time for one required inoperable LPSW pump on a temporary basis. The U.S.
Nuclear Regulatory Commission (NRC) staff has determined that additional information is needed as provided in the Enclosure. During the March 4, 2022, audit closure call, we discussed the requests for additional information to ensure mutual understanding. As discussed on the call, please respond within 45 days of the date of this e-mail.
Please note that the NRC staffs review is continuing and further requests for information may be developed. If you have any questions, please contact Shawn Williams at 301-415-1009 or via e-mail at Shawn.Williams@nrc.gov.
Sincerely, Shawn Williams, Senior Project Manager Plant Licensing Branch, II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287
Enclosure:
Request for Additional Information cc w/encl: Listserv
REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST TECHNICAL SPECIFICATION 3.7.7, LOW PRESSURE SERVICE WATER (LPSW) SYSTEM TO EXTEND THE COMPLETION TIME FOR ONE REQUIRED INOPERABLE LPSW PUMP OCONEE NUCLEAR STATION UNITS 1, 2 AND 3 DUKE ENERGY CAROLINAS, LLC DOCKET NOS. 50-269, 50-270 AND 50-287 By [[letter::RA-21-0005, Application to Revise Technical Specification 3.7.7, Low Pressure Service Water (LPSW) System, to Extend the Completion Time for One Required Inoperable LPSW Pump on a Temporary Basis|letter dated September 2, 2021]] (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21245A210) Duke Energy Carolinas, LLC (Duke Energy), submitted a license amendment request (LAR) to revise Technical Specification (TS) 3.7.7, Low Pressure Service Water (LPSW) System, to extend the Completion Time for one required inoperable LPSW Pump on a temporary basis for Oconee Nuclear Station (ONS), Units 1, 2, and 3.
Specifically, the proposed change would add a Note modifying the Completion Time (CT) associated with TS 3.7.7, Condition A, Required Action A.1, to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br /> during ONS Unit 2, Refuel 31 (fall 2023) to allow for the tie-in and testing of an alternate suction source to the shared Unit 1 and Unit 2 A and B LPSW pumps. In its submittal, Duke Energy stated that the alternate suction source to the shared Unit 1 and Unit 2 A and B LPSW pumps is needed to permit draining of the Condenser Circulating Water (CCW) System cross-connect header for the replacement of three CCW valves.
The U.S. Nuclear Regulatory Commission (NRC) staff has determined that an audit was needed to review the proposed LAR. By letter dated January 6, 2022 (ADAMS Accession No. ML21362A753), the NRC staff issued the Audit Plan. On February 10, 2022, the NRC staff conducted an audit in accordance with the Audit Plan.
The NRC staff has determined that following additional information is needed to complete our review.
Technical Specifications Branch (STSB)
Regulatory Basis Title 10 of the Code of Federal Regulations (10 CFR) Part 50 Section 50.36(c)(2)(i) states, in part:
Limiting conditions for operation [LCO] are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.
[Emphasis added]
Issue The information requested below is necessary to understand if the extension of time permitted to follow remedial action when the LCO is not met is acceptable. The licensees amendment request provides an evaluation justifying the licensees ability to perform the modification safely with only the Unit 1 and 2 LPSW C pump operable. Sections 3.2 and 3.3 of the licensees LAR show that its safety basis and risk assessment rely significantly on utilizing a capability to cross-connect the Unit 3 LPSW pumps to supply Unit 1s loads should the LPSW C pump fail during the Unit 1 and 2 LPSW modification. As stated in Section 3.2 of their LAR, This cross connect is the ACTION/EXPECTED RESPONSE for a loss of Unit 1 and 2 LPSW pumps in ONS procedure AP/1/A/1700/024, Loss of LPSW. As described, the cross-connect method would be used preferentially over natural circulation in the steam generators to remove decay heat.
The Oconee Tech Spec LCO 3.7.7 requires the LPSW to be operable. More specifically, it requires Units 1 and 2 LPSW to be operable and Unit 3 LPSW to be operable. The reason Units 1 and 2 are called out together in the LPSW LCO is because they are licensed to share their LPSW systems. Unit 3 appears to have a different design basis requirement and is not licensed to share with Units 1 and 2 in the same manner. Oconee TS Bases for LCO 3.7.7, states, The LPSW system for Unit 1 and Unit 2 is shared and consists of three LPSW pumps which can supply multiple combinations of pathways to supply required components. The LPSW system for Unit 3 consists of two LPSW pumps which can supply multiple combinations of pathways to supply required components.
The Conditions in LCO 3.7.7 define Required Actions to be taken in the event of a loss of one required LPSW pump for Units 1 and 2 (restore the required LPSW pump within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />).
Based on the above, the NRC staff requests the following information:
STSB RAI No. 1 LCO 3.7.7 has no Condition and associated Required Actions for loss of all LPSW pumps in Units 1 or 2. Please explain if this means during a loss of all LPSW situation in Units 1 or 2, the plant would be in LCO 3.0.3, that states:
When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. [Emphasis added]
STSB RAI No. 2 LCO 3.7.7 does not provide any Condition or Required Actions that allow Unit 3 to share its LPSW pumps with Units 1 or 2. Please explain if this means when Unit 3 is cross connected to Unit 1 or 2, Unit 3 would be in LCO 3.0.3 while sharing LPSW between Unit 3 and the other two units.
STSB RAI No. 3 It appears that the loss of all LPSW in Unit 1 would be a beyond-design-basis event (i.e., it is not analyzed in the Oconee FSAR [Final Safety Analysis Report]). The licensee has developed Emergency Operating Procedures to provide operators with a procedure to mitigate this event.
As stated in Section 3.2 of the LAR, the procedure directs the operators to cross-connect Unit 3 LPSW to provide the safety-related loads in Unit 1. Cross-connecting Units 1 and 3 does not appear to have a corresponding Condition in the LPSW LCO. Please explain if the use of the cross-connect would place one or both of the units in an unanalyzed condition.
STSB RAI No. 4 Since the licensee would rely on the LPSW cross-connect to mitigate a loss of all LPSW on Unit 1 and 2, please describe what compensatory measures would be taken to ensure the availability of the Unit 3 LPSW trains (e.g., protect both Unit 3 LPSW trains and required equipment during the Unit 1 and 2 modification window). In addition, to minimize a loss of all LPSW condition, please describe what compensatory measures would be taken to ensure the Unit 1 and 2 pumps maintains operability.
STSB RAI No. 5 Please describe what compensatory measures would be taken to ensure the availability of sufficient qualified operators to be able to simultaneously shutdown Units 1 and 3 should they both end up in LCO 3.0.3 or in the postulated scenario of a loss-of-coolant accident (LOCA)/loss of offsite power in one unit and an orderly shutdown in the other.
STSB RAI No. 6 The requested approval of the extended CT is solely for the completion of the modifications to the Unit 1 and 2 LPSW system. Please confirm any compensatory measures that will be taken to ensure configuration control in parallel with or subsequent to the tie in testing of an alternate suction path that could render LPSW inoperable and inappropriately extend the completion time beyond the time necessary to complete the modification being evaluated.
Mechanical Engineering and Inservice Testing Branch (EMIB)
During an audit, the licensee described the temporary alternate LPSW suction source to be installed to support the planned CCW valve replacement activity. The licensee stated that two butterfly valves will be installed in the temporary alternate LPSW suction source. The licensee stated that one of those butterfly valves will remain installed after the alternate LPSW suction source is removed following completion of the valve replacement activity. The licensee also discussed the functions of the LPSW-1095 valve, that could provide LPSW from Unit 3 to Unit 1 and 2 if the LPSW C pump failed, but was not able at the time, to discuss the design and performance of this valve. As part of the audit, the licensee subsequently made available plant documentation related to the valve type and size, function, and testing of LPSW-1095.
EMIB RAI No. 1 Please provide the following information regarding the two butterfly valves to be part of the temporary alternate LPSW suction source: (a) the functions of the two butterfly valves; (b) testing activities to demonstrate the capability of the two butterfly valves to perform its functions; and (c) the applicability of the two butterfly valves to the Inservice Testing (IST)
Program and the basis for that applicability.
EMIB RAI No. 2 Please provide the following information regarding the permanent butterfly valve that will remain installed after the temporary alternate LPSW suction source is removed: (a) the functions of the butterfly valve; (b) testing activities to demonstrate the capability of the butterfly valve to perform its functions; and (c) the applicability of the butterfly valve to the IST Program and the basis for that applicability.
EMIB RAI No. 3 Please provide the following information regarding the LPSW-1095 valve assembly. (a) type of valve (including disk) and actuator; (b) valve operating method in the open and close directions (such as electrical, manual handwheel, or actuator gearbox handwheel), (c) stroke time of the valve; (d) operating history of the valve (such as performance challenges, pressure locking, and thermal binding); (e) safety-related functions and high safety significant functions that are not classified as safety-related; (f) application of the IST requirements for this valve specified in 10 CFR 50.55a including the applicable provisions of the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants, Division 1, OM Code:
Section IST (OM Code); and (g) periodic assessment of the capability of this valve to perform its functions as credited at Oconee.
Nuclear Systems Performance Branch (SNSB)
Regulatory Basis Updated Final Safety Analysis Report (UFSAR), Section 3.1.6, Criterion 6 - Reactor Core Design (Category A), states, in part:
The reactor core shall be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all off-site power.
UFSAR, Section 3.1.10, Criterion 10 - Containment (Category A), states, in part:
Containment shall be provided. The containment structure shall be designed to sustain the initial effects of gross equipment failures, such as a large coolant
boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain for as long as the situation requires the functional capability to protect the public.
UFSAR, Section 3.1.41, Criterion 41 - Engineered Safety Features Performance Capability (Category A), states, in part:
Engineered safety features such as Emergency Core Cooling and Containment Heat Removal Systems shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety function. As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component.
UFSAR, Section 3.1.49, Criterion 49 - Containment Design Basis (Category A), states, in part:
The containment structure, including access openings and penetrations, and any necessary containment heat removal systems shall be designed so that the containment structure can accommodate without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a loss-of-coolant accident, including a considerable margin for effects from metal-water or other chemical reactions that could occur as a consequence of failure of Emergency Core Cooling Systems.
UFSAR, Section 3.1.52, Criterion 52 - Containment Heat Removal Systems (Category A),
states, in part:
Where active heat removal systems are needed under accident conditions to prevent exceeding containment design pressure, at least two systems, preferably of different principles, each with full capacity shall be provided.
SNSB RAI No. 1 Section 2.2 of the LAR states that the LPSW system supplies cooling water to the safety-related reactor building cooling units (RBCUs) system and the low-pressure injection (LPI) system coolers. These systems are used to mitigate the consequences of a LOCA.
In a scenario when only LPSW C pump is available while Unit 1 is in LCO 3.7.7 for up to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />, if a worst case design basis accident (DBA) occurs (the worst case design basis accident involves a LOCA/loss of offsite power with a loss of instrument air as described in UFSAR Section 9.2.2.2.3, page 9.2-7 last paragraph) and if the LPSW C pumps fails, a beyond design basis (defense-in-depth) capability to cross-connect the Unit 3 LPSW pump to supply Unit 1 safety related RBCU and LPI system loads will be used. Since the cross-connect is a beyond-design-basis capability proposed to be used for mitigating a DBA, the NRC staff request the following information regarding the capability of the Unit 3 to Unit 1 cross-connect.
(a) Assuming the LPSW C pump fails at any time from 0 to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />, please describe how much time it takes to establish the required LPSW flow to Unit 1 safety-related systems from the cross-connect.
(b) During the period in response to (a) containment cooling would stop completely. The UFSAR analysis of record (AOR) Figures 6-36 and 6-37 appears to indicate containment pressure and temperature profiles may reverse. With the same analysis inputs and assumptions as in the AOR, provide the containment analysis pressure and temperature profiles, peak pressure, and vapor temperature for the worst-case assuming failure of the LPSW C pump at any time from 0 to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />.
(c) Confirm that the analysis in (b) shows that the containment peak pressure and peak vapor temperature are bounded by the AOR values in UFSAR Figures 6-36 and 6-37, respectively.
(d) In case the containment peak pressure exceeds its AOR value, confirm that it is bounded by the TS Integrated Leak Rate Test (ILRT) Pressure, and the containment design pressure.
(e) In case the containment peak vapor temperature exceeds its AOR value, confirm that
- it remains bounded by the AOR equipment qualification temperature profile.
- the peak containment wall temperature based on the peak vapor temperature remains bounded by the AOR containment structural design temperature.
(f) Due to the absence of LPSW flow to RBCU and LPI coolers for the period in (a) during anytime between 0 to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />, show that the worst-case sump temperature profile remains bounded by the AOR temperature profile. If it is not bounded, please provide the following:
I. Revised sump temperature profile.
II. Available net positive suction head (NPSH) profile(s) for the pumps that draw water from the sump during the LOCA recirculation phase.
III. Minimum NPSH margin for the pumps in (ii) and the containment accident pressure (CAP) below the vapor pressure at sump temperature used in calculating this margin.
SNSB RAI No. 2 Please provide the impact on the analysis results of each of the transient and accident cases listed in UFSAR Table 15-32, due to the unavailability of LPI system coolers for the temporary period in response to SNSB-RAI 1(a) from 0 to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />.
SNSB RAI No. 3 Considering a worst case DBA in Unit 1 during which the cross-connect is supplying cooling water to Unit 1 from one of the two Unit 3 LPSW pumps, in conjunction with loss of offsite power (LOOP) in Unit 3, when one LPSW pump is unavailable due to being cross-connected, discuss how Oconee will be able to ensure that adequate decay heat removal capability will be available for Unit 3 to bring it to Mode 5.
Containment and Plant Systems Branch (SCPB)
Regulatory Basis The regulations in 10 CFR 50.36(c)(2) requires that TSs contain LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the LCO can be met. Typically, the TSs require restoration of equipment in a timeframe commensurate with its safety significance, along with other engineering considerations.
The regulations in 10 CFR 50.36(b) states, in part, The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto. In determining whether the proposed TS remedial actions should be granted, the Commission will apply the reasonable assurance standards of 10 CFR 50.40(a) and 50.57(a)(3).
The regulations in 10 CFR 50.65(4) states:
Before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of the assessment may be limited to structures, systems, and components that a risk-informed evaluation process has shown to be significant to public health and safety.
Issue Section 3.2 of the LAR states the following:
The required window to complete the tie-in and perform a functional test of the alternate suction source is projected to require 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />, which exceeds the TS 3.7.7 Completion Time for Required Action A.1 of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Although operability of the single Units 1 and 2 C LPSW Pump can provide for the specified safety function of the system for Unit 1, ONS has the capability to procedurally cross connect the ONS Unit 3 LPSW pumps to the ONS Unit 1 and 2 LPSW header by opening valve LPSW-1095 should the C LPSW Pump become inoperable.
During an audit, the licensee stated that the above cross connect would take about 15 minutes to complete, which includes dispatch and local task time.
As stated in UFSAR Section 9.2.2, Cooling Water Systems, LPSW system provides cooling water to reactor building cooling units, decay heat removal coolers, high pressure injection pump motor bearing coolers, and motor-driven emergency feedwater pump motor air coolers.
The reactor building cooling units provide containment heat removal and decay heat removal coolers provide cooling to low pressure injection in reactor building emergency sump
recirculation mode operation of low-pressure injection system following postulated LOCAs as discussed in UFSAR Chapter 6. Of the transients and accidents analyzed in UFSAR Chapter 15, steam generator tube rupture and small-break LOCA credit high pressure injection system for inventory control during the initial phase and steam line break accident credits motor-driven emergency feedwater system.
SCPB RAI No. 1 To determine the increase in risk that may result from extending TS 3.7.7, Condition A, Required Action A.1 to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />, please provided responses to the following questions related to the capability of the Unit 3 LPSW pumps to compensate for a scenario where a transient or an accident occurs and LPSW C pump fails:
(a) Explain the methods (i.e., engineering judgement, flow calculations, or flow measurements during testing) used to establish that the LPSW cross-connect from Unit 3 to Unit 1 can provide adequate flow for functions necessary to mitigate accidents or shutdown Unit 1. Discuss the ability to support the specific LPSW safety functions (i.e., cooling of high pressure injection pumps, emergency feedwater pump motor cooling, reactor building cooling units, and low-pressure injection system coolers) on a best estimate basis, consistent with the risk management program established to meet 10 CFR 50.65(a)(4).
(b) Explain the dependency of the high pressure injection pumps and EFW pump motors on LPSW cooling considering the time to cross-connect Unit 3 LPSW.