L-23-086, Response to Request for Additional Information for Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair
| ML23063A144 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/04/2023 |
| From: | Blair B Energy Harbor Nuclear Corp |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-23-086, EPID L-2023-LLA-0027) | |
| Download: ML23063A144 (1) | |
Text
Energy Harbor Nuclear Corp.
Beaver Valley Power Station P. O. Box 4 Shippingport, PA 15077 Barry N. Blair 724-682-5234 Site Vice President, Beaver Valley Nuclear March 4, 2023 L-23-086 10 CFR 50.90 10 CFR 50.91 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Response to Request for Additional Information for Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair (L-2023-LLA-0027)
By letter dated March 1, 2023 (Accession No. ML23060A018) and pursuant to 10 CFR 50.90, Energy Harbor Nuclear Corp. requested an amendment to the facility operating license for Beaver Valley Power Station, Unit No. 1 (BVPS-1). The proposed change would revise Technical Specification (TS) 3.5.2, ECCS - Operating, Limiting Condition for Operation (LCO) 3.5.2, to add a note (Note 4) allowing a one-time use of an alternate manual flow path to support repair of a leak. The use of the note would expire on April 7, 2023, at 2400 eastern daylight time (EDT). The one-time configuration addressed by the note allows for on-line repair of the leak.
The Nuclear Regulatory Commission (NRC) staff determined that additional information is needed to complete the review of the requested amendment. By electronic mail dated March 3, 2023, the NRC staff issued a request for additional information (RAI) to support the review. The Energy Harbor Nuclear Corp. RAI response is attached. To support one of the responses, a site procedure is also enclosed.
There is no change to the conclusion of the no significant hazards consideration.
Beaver Valley Power Station, Unit No. 1 L-23-086 Page 2 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 March 4, 2023.
Barry N. Blair
Attachment:
Response to Request for Additional Information
Enclosure:
1OM-53A.1.FR-C.1 (ISS3), Response To Inadequate Core Cooling cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative
Attachment L-23-086 Response to Request for Additional Information Page 1 of 16 By letter dated March 1, 2023 (Accession No. ML23060A018) and pursuant to 10 CFR 50.90, Energy Harbor Nuclear Corp. requested an amendment to the facility operating license for Beaver Valley Power Station, Unit No. 1 (BVPS-1). The proposed change would revise Technical Specification (TS) 3.5.2, ECCS - Operating, Limiting Condition for Operation (LCO) 3.5.2, to add a note (Note 4) allowing a one-time use of an alternate manual flow path to support repair of a leak. The use of the note would expire on April 7, 2023, at 2400 eastern daylight time (EDT). By electronic mail dated March 3, 2023, the Nuclear Regulatory Commission (NRC) staff issued a request for additional information (RAI) to support the review. The requested information is provided below. The request for additional information (RAI) is presented in bold font, followed by the Energy Harbor Nuclear Corp. response.
RAI-1
Emergency Core Cooling System 10 CFR 50 Appendix A General Design Criteria for Nuclear Power Plants Criterion 35 Emergency Core Cooling states:
A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.
Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
The proposed actions of this license amendment (Attachment 1 Section 4.1) request do not comply with the single failure criterion of GDC [General Design Criteria] 35, specifically if that single failure is MOV-1SI-836 failing to open, high head safety injection will not have an injection path into the reactor on a valid safety injection signal.
Attachment L-23-086 Page 2 of 16 Please provide a technical justification for maintaining the single failure criterion in the safety evaluation; or needs sufficient justification that the probability of the single failure is not sufficiently credible. Specifically,
- a. What additional risk management actions could be performed to provide additional assurance MOV-1SI-836 is functioning properly? For example, could you perform a static stroke test of the valve prior to hanging the clearance on MOV-1SI-867C and MOV-1SI-867D.
Response
The valve has no known deficiencies and has stroked properly on the last performance during the 2022 refueling outage. It is not feasible to perform a static stroke test of MOV-1SI-836 because that would result in a safety injection. The valve does get stroked in a surveillance and is monitored as part of the monthly system lineup surveillance. To address the potential for not stroking, an operator is to be briefed to deenergize and manually stroke the MOV. This is part of the Emergency Operating Procedure (EOP) developed for this contingency.
- b. Please describe any equipment history issues for MOV-1SI-836 and any associated valves in the same IST [In-service Testing] group to justify the reliability estimates of MOV-1SI-836 in Sensitivity Case 2 (Attachment 4 page 23).
Response
The IST does not group this valve with any other motor-operated valves (MOVs), but rather diagnostic tests them all individually. MOV-1SI-836 was repacked in 2004 due to boric acid leakage. Since that time, work that may affect stroke times is identified below.
October 18, 2004 Diagnostic Testing November 10, 2004 Repack March 21, 2006 Retorque packing September 24, 2007 Diagnostic testing October 2, 2010 Lube Motor May 5, 2012 Lube Motor October 15, 2016 Diagnostic Testing April 20, 2021 Diagnostic Testing Data dating back to 1999 was reviewed and shows no conditions that affected stroking of the valve. Sensitivity case 2 is presenting a more realistic failure rate for the operator failing to open MOV-1SI-836 in the control room due to having a dedicated operator who was just briefed on the evolution. The failure rate of the valve itself not opening is a separate basic event, which was not changed from its nominal value.
Attachment L-23-086 Page 3 of 16
- c. What additional risk management actions could be performed to provide a recovery contingency for HHSI [high head safety injection] against MOV-1SI-836 failing to open, even if the actions cannot be modeled in PRA
[probabilistic risk assessment]?
Response
It is correct to state that the planned configuration does not meet GDC 35 by itself; however, the execution of this activity is consistent with applicable regulatory guidance and normal industry practices. Dating back to Generic Letter (GL) 80-30, there is an acknowledgement of a change to how single failure criterion is applied when TS Actions are entered.
When the required redundancy is not maintained, either due to equipment failure or maintenance outage, action is required, within a specified time, to change the operating mode of the plant to place it in a safe condition.
The specified time to take action, usually called the equipment out of service time, is a temporary relaxation of the single failure criterion, which, consistent with overall system reliability considerations, provides a limited time to fix equipment or otherwise make it OPERABLE. If equipment can be returned to OPERABLE status within the specified time, plant shutdown is not required.
Similarly, in more modern guidance, the NRC staff affirmed in a 2007 letter between the NRC Technical Specification Branch Chief and the Pressurized Water Reactor Owners Group (Accession No. ML070240309) that the Completion Time in Technical Specifications represents a temporary relaxation of the single failure criterion.
The implementation of the maintenance that is supported by this license amendment request (LAR) will require declaring LCO 3.5.2 not met, the B train of Emergency Core Cooling System (ECCS) being declared inoperable, and 3.5.2 Action A.1 will be entered. Therefore, this configuration is consistent with GDC 35 when considered in the context of the relationship between the General Design Criteria and Technical Specifications. The safety function will be preserved, but it does not have required redundancy so a limited time to fix equipment is provided.
The mitigation to the loss of all HHSI is performed in the EOP network. On a small break loss of coolant accident where reactor coolant system (RCS) pressure remains above the shutoff head of the low head safety injection (LHSI) pumps, the RCS is cooled by the steam generators and depressurized to allow the LHSI system to inject water. This is handled by the EOP network through the use of the functional restoration procedure for response to inadequate core cooling, a copy of which is enclosed with this submittal.
Attachment L-23-086 Page 4 of 16
RAI-2
Fire PRA Model Results [Human Factors Review]
Regulatory Basis:
The NRC staff reviews the human performance aspects of licensing action requests utilizing guidance in NUREG-1764, Revision 1, Guidance for the Review of Changes to Human Actions (ML072640413). NUREG-1764 describes human factors reviews as Level I (high risk) or Level II (medium risk) with the possibility of reduction to a Level III (low risk) review, if appropriate. The licensees submittal dated March 1, 2023, proposes new manual operator actions associated with a risk-significant system. Per NUREG Section 1764, Rev. 1, Section 2.4, Screening Process for Non-Risk-Informed Change Requests, and Table A.2, Generic PWR Human Actions That Are Risk-Important, the proposed actions are considered potentially risk significant and a Level II human factors review is appropriate. The information requested below is required to enable the NRC staff to perform a Level II human factors review on the proposed manual actions. , Section 4.2, response to item number 3, states the following, in part:
During the period the note is invoked, the ECCS will remain capable of mitigating the consequences of a design basis event such as a loss-of-coolant accident. In addition, simulator runs have validated that the manual action can be reliably performed in the necessary timeframe to meet the accident analysis.
- a. Provide a description of the referenced simulator runs.
Response
With the simulator set up in Mode 1 at 100% power, the boron injection tank (BIT) isolation valves MOV-1SI-867C and MOV-1SI-867D were closed and deenergized to simulate the required clearance. An operator was stationed at the bench board in the area of MOV-1SI-836, HHSI pump discharge to RCS cold legs. A 1000 gallon per minute (gpm) loss-of-coolant accident (LOCA) was inserted. No operator action was taken until the automatic safety injection (SI) setpoint of 1845 pounds per square inch gauge (psig) in the pressurizer was exceeded as indicated by Annunciator A5-31, Pressurizer Press Low Reactor Trip and SI. When Annunciator A5-31 alarmed, action from the dedicated operator was initiated to open MOV-1SI-836 from the control room to provide flow to the core.
- b. Describe the analysis of the time required vs. time available to open the MOV-1SI-836 and fulfill the Safety Injection function.
Attachment L-23-086 Page 5 of 16
Response
As stated in Section 14.3.1 of the Updated Final Safety Analysis Report (UFSAR), the accident analysis assumes the HHSI system will deliver water to the RCS 27 seconds after the generation of a SI signal. The 27 second delay includes time required for signal processing, diesel startup and loading of the safety injection pumps onto the emergency buses, as well as the pump acceleration and valve delays. This scenario was validated on the simulator with manual action of opening the valve from the control room, and the time was determined to be 12 seconds.
Should the valve not open from the control room, an operator, who will be stationed in the field, will operate the valve locally. Local operation time is discussed in more detail below but is assumed to take three minutes. This compares favorably to the PRA analyses for a small-break LOCA as well as a medium-break LOCA. The small-break LOCA requires HHSI within 58 minutes and is based upon the latest time injection flow could be initiated that would prevent core damage. The most limiting PRA timing is for the medium-break LOCA, which requires HHSI within 21 minutes and is based upon the latest time injection flow could be initiated that would prevent core damage. HHSI is not credited for a large-break LOCA based on modular accident analysis program (MAAP) analyses performed to support the PRA.
- c. Provide the manual actuation times achieved by the control room operator during the simulator runs.
Response
The manual actuation times achieved by the control room operator during the simulator runs are as follows:
T-0: safety injection actuates T+5 seconds: MOV-1SI-836 is taken to open T+12 seconds: Safety Injection (SI) flow > 400 gpm verified on FI-1SI-940
- d. Provide the manual actuation times achieved by the local field operator during the simulator runs.
Response
The field operator will be stationed in Safeguards Building (West Cable Vault) at MCC1-E5 and will be in constant communication with the control room. The field operator will have been pre-briefed for entry into A Penetrations room and be signed onto the appropriate radiation work permit (RWP). The operator will have Standing Order 23-003 and the EOP in hand.
Attachment L-23-086 Page 6 of 16 Because the field operator is stationed in the vicinity of the power supply one minute is allotted to open the associated breaker.
An additional minute is allotted to safely travel from MCC1-E5 to the location of MOV-1SI-836.
Previous simulation for NFPA-805 validation credited 30 seconds to manually operate MOV-1SI-836. Accessibility to the valve was re-verified and validated.
Total time to locally operate MOV-1SI-836 after direction is given is estimated to be 3 minutes.
- e. Describe the alerts/cues to perform the manual action that are provided to the control room operator.
Response
Standing Order 23-003 will direct the dedicated control room operator to perform the EOP written for this specific activity. The standing order identifies the guidance is effective upon isolation of the BIT to affect the removal of RV-1SI-857 and will remain in effect until the BIT flowpath is restored to normal system alignment. The specific alerts/cues from the guidance include the following:
Upon receipt of an automatic safety injection or actuation of a manual safety injection, and without delay, the dedicated reactor operator assigned will open MOV-1SI-836 in accordance with the EOP.
When MOV-1SI-836 is open, the dedicated control room operator will verify HHSI flow on FI-1SI-940.
EOP guidance for verification of and subsequent isolation of the HHSI flowpath will be used throughout the EOP network while in the off-normal configuration.
Multiple Control Room annunciators and indicators indicate the need for safety injection including:
A5-29, Containment Pressure High Reactor Trip & SI A5-31, Pressurizer Press Low Reactor Trip & SI A5-13, Low STM Line Press Reactor Trip SI & STM Line ISOL Negative RCS Pressure and Pressurizer Level Trends
- f. Describe the alerts/cues to perform the manual action that are provided to the operator in the field location.
Attachment L-23-086 Page 7 of 16
Response
Constant communication will be established with the control room on an open, dedicated phone line with the operator in the West Cable Vault. The control room will perform a once per 15-minute check-in with the local operator to ensure alertness and readiness to respond.
If the crew is unsuccessful opening MOV-1SI-836 from the control room. The EOP Step 1 Response Not Obtained Column will direct the designated operator in the field to deenergize and locally open MOV-1SI-836.
- g. Describe the tasks involved in local field actuation of MOV-1SI-836 and the method by which the MOV-1SI-836 will be actuated.
Response
The field operator removes power from the valve motor by opening the supply breaker in the West Cable Vault, Safeguards 735. Then the field operator enters A Penetrations Room, Safeguards 722, and declutches the motor operator. The valve is manually opened using the handwheel.
- h. Describe the tasks involved in control room actuation of MOV-1SI-836.
Response
MOV-1SI-836 is a manually opened MOV controlled by a single control switch. No other tasks are required to open the valve beyond operation of the control switch on the bench board panel in the control room. The requirement to operate the control switch is clearly defined and indicated in the control room.
- i. Describe the required mitigation of any environmental impacts for the local field operator location.
Response
The dedicated operator will be stationed in the West Cable Vault room, which is a low dose, low noise area. It maintains a reasonable room temperature and is lightly transited. It is an ideal location that is free of environmental impact. This allows the operator to stand ready to perform the required actions. Constant communication will be established with the control room on an open dedicated phone line with the operator in the West Cable Vault. The control room will perform once per 15-minute check-ins with the local operator to ensure alertness and readiness to respond.
- j. Describe the integration and command and control/communications for the operating staff designated to perform the manual actions.
Attachment L-23-086 Page 8 of 16
Response
At the beginning of the shift, the crew will have a pre-evolution brief. At that brief, required roles and responsibilities will be covered. Constant communication requirements will be covered. The crew will brief the standing order and EOP. The Unit Assistant Operations Manager will attend the brief and ensure all participants display complete understanding of the standing order, EOP, required communications, and so forth.
- k. Describe the procedures/instructions that will direct the manual actions both in the control room and at the MOV-1SI-836 field location.
Response
As previously mentioned, a standing order is to be used. Standing Order 23-003 will provide guidance for ECCS operation with the BIT flowpath isolated as described in response to RAI-2 (e) above.
Also as previously mentioned, an EOP procedure is to be issued. The procedure is titled Alternate HHSI Flowpath Verification and provides a list of actions required to establish and verify an alternate SI flowpath through MOV-1SI-836 while normal HHSI flowpath through MOV-1SI-867C and MOV-1SI-867D is inoperable.
- l. Describe the training that has been or will be provided to operators responsible for implementing operator actions both in the control room and at the MOV-1SI-836 field location.
Response
At the beginning of each shift, the crew will have a pre-evolution brief. At that brief, required roles and responsibilities will be covered. The tasks are considered normal operator functions and require no additional training beyond briefing.
- m. Describe the measures to be put in place to ensure that the control room operator and local field operator will remain alert and capable of responding to a Safety Injection signal for the duration the alternate flow path configuration.
Response
The designated control room operator will remain in the red carpeted controls area and will have no other duties. Upon receipt of an SI signal, the operator will immediately open MOV-1SI-836 and verify flow on FI-1SI-940. The operator will be pre-briefed on the expectations. The operator will not leave the controls area without being relieved by
Attachment L-23-086 Page 9 of 16 another qualified operator with no other duties who has also been pre-briefed on the expectations.
The designated field operator will be in constant communication with the control room operator. The control room will perform a once per 15-minute check-in with the local operator to ensure alertness and readiness to respond. The operator will not leave the area without being relieved by another qualified operator with no other duties who has also been pre-briefed on the expectations.
RAI-3
Fire PRA Model Results As shown in the risk results, it appears over 90% of the increase in risk is from the fire PRA scenarios. The LAR further explains that the risk increase associated with the proposed plant configuration is dominated by fire scenarios in which a LOCA may result, whether by spurious opening of a PORV [power operated relief valve], a valid PORV demand with failure to properly re-close, fire-induced failure of PORVs to open resulting in challenging a primary safety valve which fails to reclose, failure of RCP seal injection and shutdown seals (SDS) with a resultant RCP seal LOCA, etc.
The LAR states the Fire PRA model was upgraded to the state-of-the art in order to support the NFPA 805 fire protection licensing basis, issued in letter dated January 22, 2018. It also indicates that the PRA model used as the basis for the risk assessment provided in the LAR is the PRA average maintenance model of record issued on January 5, 2023.
Response
There were changes since the approval of NFPA-805 specific to the fire model. The changes listed below were classified as upgrades and were reviewed in a focused scope peer review of the Fire PRA (FPRA) performed in October 2017. This peer review was performed using the processes defined in Nuclear Energy Institute (NEI)
DC hot short duration (in accordance with NUREG/CR-7150, Volume 2)
Obstructed plume (in accordance with NUREG-2178, Volume 1)
Use of Fire Dynamics Simulator (FDS)
Incipient Detection credit (in accordance with NUREG-2180)
Supporting Requirements CF-A1, CF-A2, CF-B1, FSS-C4, FSS-C7, FSS-D1, FSS-D2, FSS-D3, FSS-D4, FSS-D7, FSS-D8, FSS-D10, FSS-H3, FSS-H4, and FSS-H10 were
Attachment L-23-086 Page 10 of 16 reviewed to verify proper implementation of the listed upgrades, and all were determined to be Met at Capability Category II or higher. There were no Findings. One suggestion level Finding and Observation (F&O) was assigned to FSS-C7 and FSS-D7, which reads:
Issue:
FENOC reports 2701.620000017 and 2701.620000.109 do not address the use of incipient detection at Beaver Valley.
Basis for Significance:
Incipient detection is credited in the BV Fire PRAs but is not addressed in all applicable reports. Because the system has not been installed very long, there is no impact on the PRA inputs or current peer review conclusions.
Possible Resolution:
Update 2701.620-000-017 and 2701.620-000.109 as appropriate to address the use of incipient detection at Beaver Valley.
Closure:
Both reports were updated to address the use of incipient detection.
2701-620-000-017 is NFPA 805 Fire PRA Task 5.11A Fire Detection and Suppression System Dependency. 2701-620-000-109 is Evaluation of Fire Protection System Unavailability. This Suggestion F&O was focused on updating program documentation to include the newly installed incipient detection system. The program documents were updated to account for incipient detection. This Suggestion has no impact on this risk assessment because it was primarily a documentation issue that has since been resolved.
Notable maintenance changes to the fire PRA model are listed below:
Fire modeling refinements were made to four ignition sources in 1-NS-1 (Normal Switchgear), to refine scenarios using specific source-to-target measurements.
Additionally, limited credit for instrument air, crediting only the diesel driven air compressor, was added to fires in 1-NS-1 based on drawing reviews and a confirmatory walkdown to verify no instrument air piping exists in this compartment. This allows some credit for RCP Thermal Barrier cooling (dependent upon instrument air) for fires in 1-NS-1, and reduces the likelihood of reactor coolant pump (RCP) seal LOCA due to fire in this compartment.
Based on PWROG-14001-P R1 the RCP SDS modeling was updated to account for the effect of the SDS temperature qualification related to asymmetric cooldown.
Specifically, the PRA model was updated to ensure that all three steam generators must be fed for all accidents demanding use of the shutdown seals in order to avoid
Attachment L-23-086 Page 11 of 16 exceeding the SDS qualification temperature. If all three steam generators are not fed, the shutdown seals are assumed to fail.
Minor updates were made to cable selection, cable routing, and equipment counts for ignition frequency bins, to account for plant changes.
Beyond these specific changes, standard periodic PRA model updates have been performed in accordance with Energy Harbor Business Practice NOBP-CC-6001, Probabilistic Risk Assessment Model Management.
- b. Provide a discussion of sources of uncertainties in the fire PRA (e.g., RCP seal models, incipient detection, etc.) that may impact the risk results and how those were considered for the risk calculation supporting this amendment.
Explain the rationale for limiting the sensitivity studies to the operator action to align the open the alternate injection flow path through MOV-1SI-836.
Response
Incipient detection is credited for numerous fire scenarios in 1-CR-4 and is modeled in accordance with NUREG-2180. This system is installed in individual cabinets in this fire compartment and provides earlier detection of fires than standard ceiling-mounted early warning detection systems. Typically, uncertainties regarding credit for incipient detection stem from use of the system in whole-room applications, but incipient detection at Beaver Valley is only credited to improve detection times in the cabinet fire scenarios where the incipient detectors are installed. Nonetheless, any remaining uncertainties related to FPRA credit for this system relative to this amendment are accounted for by the continuous fire watch to be implemented in 1-CR-4, which provides additional assurance that any fires igniting while in the proposed configuration will be promptly detected and suppressed.
RCP seals, including the Westinghouse Generation III SDS, are modeled in accordance with the guidance in PWROG-14001-P R1 and PWROG-14006-P Rev 0-B. Impact of uncertainties from the RCP seal modeling is reduced from conservatisms built into the model. The FPRA model assumes failure of instrument air for all fire scenarios (except in 1-NS-1) and subsequent failure of RCP Thermal Barrier cooling, which depends on air-operated valves to supply cooling water to the RCP Thermal Barriers. This assumption leads to a higher likelihood of RCP seal LOCA in the Fire PRA, since only seal injection remains to cool the RCP seals. Instrument air is assumed to fail due to the system being primarily constructed of brazed copper piping which cannot be ensured to hold integrity if exposed to a fire; however, a specific review was performed to verify the absence of instrument air piping in compartment 1-NS-1, allowing limited credit for instrument air (using the independent diesel driven air compressor) and RCP Thermal Barrier cooling. If such a review were to be conducted in additional fire compartments, it is likely the risk contribution of RCP seal LOCAs could be reduced to some extent.
Attachment L-23-086 Page 12 of 16 Another uncertainty directly affecting this assessment is the fact that the PRA model takes no credit for the local operator action to open MOV-1SI-836 using the handwheel.
The PRA only credits the control room action to open MOV-1SI-836 using the benchboard control switch. Therefore, any failure of the A Train emergency power supply guarantees failure of the valve in the PRA results, without accounting for the dedicated local operator stationed in the field. If credit for the local operator were added to the model, the incremental conditional core damage probability (ICCDP) and incremental conditional large early release probability (ICLERP) values would be lower than currently reported in this assessment.
As an additional measure to account for potential uncertainties in the PRA model, this amendment is requesting only half of the maximum allowable time supported by the risk assessment.
The sensitivity studies performed to support this assessment were determined based on the results of the proposed configuration as modeled in B1R8HCL1. Results of this model were reviewed, and the top risk contributors were identified to involve HHSI demands (some variety of LOCA) with failure of the alternate SI flow path (either failure of MOV-1SI-836 itself, failure of A Train emergency power to the valve, or failure of the operator action to open the valve). The risk contributors with the greatest change from the base PRA model results were also reviewed and were similarly determined to be associated with failure to provide HHSI flow through the alternate path. Given that some failure to provide HHSI flow through this alternate path dominates the risk increase in the top contributors, it was determined to focus sensitivity studies on the uncertainties involved in failure of the alternate SI flow path.
RAI-4
Top Risk Contributors and Proposed Compensatory Measures LAR Attachment 5 Section 6 provide a review of the top risk contributors per Tier 2 of RG 1.177. It identified and discussed three top risk contributors: small LOCA initiating events, fire scenario in fire compartment 1-CR-4 (Process instrumentation room) and fire scenarios in fire compartment 1-ES-1 (Train A Emergency Switchgear Room) as top risk contributors. The LAR also proposed compensatory measures associated with these risk scenarios, including continuous fire watches and no hot work permitted in the fire compartments 1-CR-4 and 1-ES-1.
Additionally, LAR Attachment 5 Section 6 states the following:
Other significant contributors which also saw a substantial risk increase due to the proposed configuration are individual fire scenarios in the Main Control Room, 1-CS-1 (cable spreading room) and 1-CV-1 (West Cable Vault).
Attachment L-23-086 Page 13 of 16
- a. Provide a summary of contribution to risk from other risk scenarios, by fire compartments.
Response
Below is a table to summarize the contribution to fire risk by fire compartments. This shows that over 97 percent of total delta core damage frequency (CDF) contribution is from Fire CDF.
Compartment CDF from Compartment CDF Compartment Delta CDF from BVPS-1 Delta CDF Percent Contribution of Fire BVPS-1 Effective from proposed repair Effective Reference PRA Model Compartment from BVPS-1 Effective Reference Fire Compartment Reference PRA Model configuration model BV1REV8 to proposed repair PRA Model BV1REV8 Baseline Total CDF to Compartment Description BV1REV8 B1R8HCL1 configuration model B1R8HCL1 proposed repair configuration model B1R8HCL1 Process Instrumentation 1-CR-4 Room 1.33E-05 6.72E-05 5.39E-05 44.29%
Emergency Switchgear 1-ES-1 Train A 2.93E-06 4.79E-05 4.50E-05 36.99%
West Cable 1-CV-1 Vault 1.55E-06 7.47E-06 5.92E-06 4.86%
Cable Spreading 1-CS-1 Room 1.76E-06 7.66E-06 5.90E-06 4.85%
3-CR-1 (MCR) Control Room 2.55E-06 6.78E-06 4.23E-06 3.47%
Normal Switchgear 1-NS-1 Room 1.67E-05 1.86E-05 1.97E-06 1.62%
Diesel Generator 1-DG-1 Cubicle A 5.29E-07 1.26E-06 7.30E-07 0.60%
Reactor Containment 1-RC-1 Building 1.36E-06 1.80E-06 4.35E-07 0.36%
Emergency Switchgear 1-ES-2 Room Train B 3.49E-07 4.40E-07 9.10E-08 0.07%
Communications Equipment and 1-CR-3 Relay Room 2.14E-06 2.19E-06 4.95E-08 0.04%
1-CV-2 East Cable Vault 1.01E-06 1.05E-06 3.98E-08 0.03%
3-ER-1 ERF Substation 4.26E-07 4.41E-07 1.47E-08 0.01%
1-TB-1 Turbine Building 2.00E-06 2.01E-06 8.93E-09 0.01%
Note: All other Fire Compartments not listed contributed to less than 0.01% total delta CDF contribution.
- b. Explain the rationale on how the risk scenarios with compensatory measures were selected. Justify that no additional compensatory measures were considered for other significant risk contributors.
Response
Referring to the table in the response in RAI-4 a, it can be shown that over 81 percent of total delta CDF contribution of the proposed repair configuration is due to fire risk in 1-CR-4 (Process instrumentation room) and 1-ES-1 (Train A Emergency Switchgear Room). As this is the majority of the delta risk contribution, it was determined that compensatory measures should remain focused on the highest risk factors. Additionally,
Attachment L-23-086 Page 14 of 16 fires in the Main Control Room, 1-CS-1 (cable spreading room) and 1-CV-1 (West Cable Vault) were shown to be the next highest delta CDF contributions although these compartments only had a few scenarios that drove the overall delta CDF for the compartment. These individual scenarios had CDF contributions between 1% and 4%
(1 in the Main Control room FMCR77, 1 in 1-CS-1 FCS1T1 and 2 in 1-CV-1 FCV160 and FCV117). These few individual scenarios have a low overall impact compared to the top compartments (1-CR-4 and 1-ES-1) for which compensatory actions are being taken.
RAI-5
Motor Operated Valve Functionality Regulatory Basis: 50.55a(b)(3)(ii) OM condition: Motor-Operated Valve (MOV) testing states in part that:
Licensees must comply with the provisions for testing MOVs in ASME OM Code, ISTC 4.2, 1995 Edition with the 1996 and 1997 Addenda, or ISTC-3500, 1998 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(iv) of this section and must establish a program to ensure that MOVs continue to be capable of performing their design basis safety functions.
In your LAR, Section 3.3, Compensatory Measures, Alternate SI Alignment with LCO 3.5.2 Note 4 Invoked, item number 3, states:
MOV-1SI-836 will remain energized and closed. MOV-1SI-836 will be available and operable, and capable of being opened manually via control switch on control room bench board. An extra, dedicated Reactor Operator will be assigned to this task as described in Section 3.1. In the event that MOV-1SI-836 fails to stroke, an extra, dedicated operator will be assigned to locally manually open MOV-1SI-836. Actions required to establish and verify the alternate SI flow path will be governed by a site procedure.
Please provide the following additional information regarding MOV-1SI-836:
- a. It is not clear how the MOV-1SI-836 valve will be opened. Will the valve be opening with the handwheel or by pushbutton?
Response
MOV-1SI-836 will be operated from the Control Room bench board by switch. If it fails to operate by switch, a dedicated operator will deenergize and manually open the valve using the handwheel.
Attachment L-23-086 Page 15 of 16
- b. The Beaver Valley Inservice Testing (IST) Program Plan dated October 10, 2017, specifies (ML17289A214) MOV-1SI-836 as a normally closed 3-inch motor-operated gate valve with required stroke-time testing open and closed every cold shutdown or refueling outage, and leak testing and remote position verification every 2 years. The IST Program Plan also indicates that diagnostic testing in the open direction is performed per ASME OM Code Case OMN-1 every 3 refueling outages. When were these tests conducted for MOV-1SI-836?
Response
Based on past good performance, leak testing is now performed on a four-refueling outage frequency per NRC approved Valve Relief Request No. 4. The most recent test was performed in October 2022.
The OMN-1 tests were last done in the spring 2016 and fall 2021 refueling outages.
- c. Has an open capability evaluation for MOV-1SI-836 been performed in accordance with 10 CFR 50.55a(b)(3)(ii)? What is the calculated capability margin? What assumptions are included in that calculation?
Response
The open capability evaluation was last performed in April 2021 using tested data input from the torque calculation. MOV-1SI-836 has high margin with no anomalies noted.
- d. Has dynamic testing of MOV-1SI-836 been performed with or without diagnostics? If diagnostics were used, what capability margin was determined from that testing?
Response
MOV-1SI-836 was dynamically tested with diagnostics in February 1995 as the GL 89-10 qualifying basis. The dynamic test results show positive margin, but the calculated margin assessment from 1995 is not readily available. MOV-1SI-836 is also stroked under dynamic conditions without diagnostics during the HHSI full flow test, which is performed each refueling outage (18 months).
- e. What type of gate valve is used in MOV-1SI-836, such as flexible wedge, solid wedge, or parallel disk? What actions have been taken to avoid pressure locking and thermal binding of MOV-1SI-836?
Attachment L-23-086 Page 16 of 16
Response
MOV-1SI-836 is a flex wedge gate valve. MOV-1SI-836 was evaluated for pressure locking and thermal binding and determined to be not susceptible. Therefore no modifications were performed.
- f. When was the last time the handwheel used to operate MOV-1SI-836 under static or dynamic conditions to demonstrate that it is properly sized and in good working condition?
Response
No record was located of the last time the MOV was operated by hand to support operations or with dynamic conditions. It was stroked satisfactorily from the bench board in October 2022 during refueling outage 28.
The MOV-1SI-836 handwheel was previously used under static conditions during the last static test and inspection in April 2021 during refueling outage 27.
- g. Do the plant procedures prohibit the use of extension bars (referred to as cheater bars) with the handwheel to manually operate Beaver Valley MOVs, including MOV-1SI-836?
Response
The plant procedure that governs the process for maintaining plant status control does not allow the use of valve wrenches or extension devices on MOVs.
No extension or cheater bars are allowed to be applied to the MOV handwheel at BVPS-1.
- h. When was the MOV-1SI-836 visually examined previously to verify that there are no indications of damage to the housing (such as cracking near the bolt holes) or stem nut (such as bronze shavings below the actuator)?
Response
MOV-1SI-836 was last visually examined for stem nut wear, such as bronze shavings below the actuator and stem nut transition time changes, in April 2021 during refueling outage 27 static testing and inspection. The inspection did not identify any bronze shavings below the actuator, and transition times had no major changes that indicate stem nut wear.
A KALSI exam of MOV-1SI-836 was also performed on 11/2/22 during the most recent refueling outage. No linear or crack like indications were observed in the examinations.
Record Type #A9.350C Beaver Valley Power Station UNIT 1 1OM-53A.1.FR-C.1(ISS3)
Response To Inadequate Core Cooling Issue 3 Revision 3 Prepared by Date Pages Issued Effective Date M. Kesow 8/1/22 1 through 17 9/1/22 Reviewed by Date Validated by Date M. P. Flynn 8/4/22 N/A PORC Meeting No. Date PAF-22-01090 PORC Not Required N/A
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 A. PURPOSE This procedure provides actions to restore core cooling.
B. SYMPTOMS AND ENTRY CONDITIONS This procedure is entered from F-0.2, Core Cooling Critical Safety Function Status Tree, on either RED condition.
C. MAJOR ACTION CATEGORIES
- 1. Establish Safety Injection Flow to the RCS.
1FRC1 8/24/2022 1 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1 Check RWST Level - GREATER THAN Go to ES-1.3, Transfer To Cold 19 FEET Leg Recirculation, Step 1.
2 Verify SI Valve Alignment - PROPER Manually align valves as EMERGENCY ALIGNMENT necessary.
3 Verify SI Flow In Both Trains Start pumps and align valves as necessary. Continue efforts to
- LHSI Flow - INDICATED from any other form of RCS injection available.
4 Check RCP Support Conditions - Try to establish support AVAILABLE conditions.
- Steam bubble in PRZR
- No. 1 seal delta-P - GREATER THAN 200 PSID
- RCP seal injection flow -
BETWEEN 6 AND 9 GPM
- Seal leakoff flow - GREATER THAN 0.2 GPM
- RCP annunciators - NONE LIT 1FRC1 8/24/2022 2 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 5 Check SI Accumulator Isolation Valve Status
- a. Power to [MOV-1SI-865A,B,C] - a. Restore power to valves:
AVAILABLE
- [MOV-1SI-865A] -
[MCC1-E5] Cub BC (West Cable Vault - 735)
- [MOV-1SI-865B] -
[MCC1-E6] Cub AZ (East Cable Vault - 735)
- [MOV-1SI-865C] -
[MCC1-E6] Cub AY (East Cable Vault - 735)
- b. [MOV-1SI-865A,B,C] - OPEN b. Open isolation valves unless closed after accumulator discharge:
- 1) Insert shorting bars into jacks for
[MOV-1SI-865A,B,C].
- 2) Open [MOV-1SI-865A,B,C].
- 3) Remove shorting bars.
6 Check Five Hottest Core Exit TCs - Go to Step 9.
LESS THAN 1200F 7 Check RVLIS Full Range Indication
- a. RCPs - NONE RUNNING a. RETURN TO procedure and step in effect.
- b. Indication - GREATER THAN 40% b. IF rising, THEN RETURN TO Step 1.
IF NOT rising, THEN GO TO Step 8.
- c. RETURN TO procedure and step in effect.
1FRC1 8/24/2022 3 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 8 Check Core Exit TCs
- a. Temperature - FIVE HOTTEST TCs a. IF dropping, THEN RETURN TO LESS THAN 719F Step 1.
IF NOT dropping, THEN GO TO Step 9.
- b. RETURN TO procedure and step in effect.
9 Check PPDWST Level - GREATER THAN Refer to Attachment 2-H for 28.0 FEET makeup.
CAUTION A faulted or ruptured SG should NOT be used in subsequent steps unless no intact SG is available.
10 Check Intact SG Levels
- a. Narrow range level - GREATER a. Maintain total feed flow THAN 31% [50% ADVERSE CNMT] greater than 370 GPM until narrow range level greater than 31% [50% ADVERSE CNMT] in at least one SG.
IF total feed flow greater than 370 GPM can NOT be established, THEN continue attempts to establish a heat sink in at least one SG and Go to Step 20.
- b. Control feed flow to maintain narrow range level between 31%
[50% ADVERSE CNMT] and 65%.
1FRC1 8/24/2022 4 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 11 Check Station Instrument Air Perform either of the following:
Header Pressure - GREATER THAN 100 PSIG
- Manually start at least one Station Air Compressor
-OR-
- Dispatch an operator to locally start [1IA-C-4],
Diesel-Driven Air Compressor IF station instrument air header pressure can NOT be restored, THEN check if an AFW Pump should be stopped. Refer to Attachment 2-S.
12 Check RCS Vent Paths
- a. Power to PRZR PORV block valves a. Restore power to valves:
- AVAILABLE
- [MOV-1RC-535] -
[MCC1-E5] Cub BE (West Cable Vault - 735)
- [MOV-1RC-536] -
[MCC1-6] Cub BC (East Cable Vault - 735)
- [MOV-1RC-537] -
[MCC1-6] Cub BD (East Cable Vault - 735)
IF any valve can NOT be closed, THEN manually close its block valve.
- c. Block valves - AT LEAST ONE c. Open block valve unless it was OPEN closed to isolate an open PORV.
- d. Reactor Coolant Vent System d. Manually close valves.
Valves - CLOSED 1FRC1 8/24/2022 5 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE
- Partial uncovering of SG tubes is acceptable in the following steps.
- After the low steamline pressure SI signal is blocked, main steamline isolation will occur if the high steam pressure rate setpoint is exceeded.
13 Depressurize All Intact SGs To 230 PSIG
- b. Check Station Instrument Air b. Manually or locally dump steam Header Pressure - GREATER THAN at maximum rate using:
100 PSIG
- SG Atm Dump Valves
-OR-
- IF no ruptured SG exists, THEN use Residual Heat Release Control Valve Refer to Attachment 2-U, Local Operation of SG Atmospheric Steam Dump Valves.
Go to Step 13.g.
(step continued next page) 1FRC1 8/24/2022 6 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 13 (continued from previous page)
- c. Check MSIVs - AT LEAST ONE OPEN c. Manually or locally dump steam at maximum rate using:
- SG Atm Dump Valves
-OR-
- IF no ruptured SG exists, THEN use Residual Heat Release Control Valve Refer to Attachment 2-U, Local Operation of SG Atmospheric Steam Dump Valves.
Go to Step 13.g.
- d. Check condenser - AVAILABLE d. Manually or Locally dump steam at maximum rate using:
- SG Atm Dump Valves
-OR-
- IF no ruptured SG exists, THEN use Residual Heat Release Control Valve Refer to Attachment 2-U, Local Operation of SG Atmospheric Steam Dump Valves.
Go to Step 13.g.
(step continued next page) 1FRC1 8/24/2022 7 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 13 (continued from previous page)
- e. Check condenser steam dump mode e. IF STM PRESS previously selector in TAVG mode. selected, THEN perform the following:
- 1) Place steam dump pressure controller in MAN
- 2) Gradually raise demand to initiate RCS cooldown
- 3) IF necessary, defeat TAVG interlock
- 4) Raise demand to obtain maximum cooldown rate
- 5) Go to Step 13.g.
- f. Dump steam to condenser at maximum rate:
- 1) Place steam dump pressure controller in MAN
- 2) Verify demand - ZERO
- 3) Place steam dump control in STM PRESS Mode
- 4) Check TAVG - GREATER THAN 4) Perform the following:
541F a) Defeat TAVG interlock
- Status light D-11, 2/3 Lo-Lo TAVG (Panel 622) - b) Gradually raise steam NOT LIT dump rate
- 5) Gradually raise steam dump c) Go to Step 13.g.
rate
- 6) As TAVG approaches 541F, 6) IF steam dumps close due defeat TAVG interlock until to TAVG interlock. THEN status light A-12, Stm Dump perform the following:
Defeat Interlock (Panel 622) - LIT a) Verify steam dump controller demand -
ZERO b) Defeat TAVG interlock c) Raise demand to obtain maximum cooldown rate (step continued next page) 1FRC1 8/24/2022 8 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 13 (continued from previous page)
IF NOT, THEN GO TO Step 20.
- AT LEAST TWO LESS THAN 410F dropping, THEN RETURN TO Step 10.
IF NOT, THEN GO TO Step 20.
- i. Stop SG depressurization.
1FRC1 8/24/2022 9 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION IF offsite power is lost after SI reset, manual action may be required to restart safeguards equipment.
14 Check If SI Accumulators Should Be Isolated
- a. At least two RCS hot leg a. Go to Step 20.
temperatures - LESS THAN 410F
- c. Insert shorting bars into jacks for [MOV-1SI-865A,B,C]
(step continued next page) 1FRC1 8/24/2022 10 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 14 (continued from previous page)
- d. Close [MOV-1SI-865A,B,C] d. Perform the following:
- 1) Reset CIA and CIB.
- 2) Verify at least one station air compressor or the diesel air compressor is RUNNING.
- 3) Verify [TV-1IA-400] -
OPEN
GREATER THAN 85 PSIG IF NOT, THEN open
[1IA-90] Instr Air Vlv To CNMT Instr Air Isol Vlv (West Cable Vault -
735).
- 5) Vent any unisolated accumulators to atmospheric pressure.
Refer to 1OM-11.4.H, Venting Safety Injection Accumulators
[1SI-TK-1A(1B)(1C)]
IF an accumulator can NOT be isolated or vented, THEN consult the TSC to determine contingency actions.
- e. Remove shorting bars 15 Stop All RCPs 1FRC1 8/24/2022 11 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 16 Depressurize All Intact SGs To Atmospheric Pressure
- a. Dump steam to condenser at a. Manually or locally dump steam maximum rate. at maximum rate from intact SGs using:
- SG Atm Dump Valves
-OR-
- IF no ruptured SG exists, THEN use Residual Heat Release Control Valve Refer to Attachment 2-U, Local Operation of SG Atmospheric Steam Dump Valves.
17 Verify SI Flow Continue efforts to establish SI flow. Try to establish flow from
-OR-IF five hottest core exit TCs
- LHSI Flow - INDICATED less than 1200F, THEN RETURN TO Step 16.
IF NOT, THEN GO TO Step 20.
18 Check Core Cooling
- a. Core exit TCs - FIVE HOTTEST a. GO TO Step 20.
TCs LESS THAN 1200F
- b. At least two RCS hot leg b. RETURN TO Step 16.
temperatures - LESS THAN 350F
- c. RVLIS full range - GREATER THAN c. RETURN TO Step 16.
64%
19 GO TO Step 16 Of E-1, "Loss Of Reactor Or Secondary Coolant" 1FRC1 8/24/2022 12 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE
- Normal conditions are desired but NOT required for starting the RCPs.
- Preference for starting RCPs under inadequate core cooling conditions is B, A, C.
20 Check If RCPs Should be Started
- a. Check five hottest core exit a. GO TO Step 21.
TCs - GREATER THAN 1200F
- b. Check if an idle RCS cooling b. Perform the following:
loop is available: 1) Reset SI, CIA and CIB.
- Narrow range SG level - 2) Verify at least one GREATER THAN 31% [50% station air compressor or ADVERSE CNMT] the diesel air compressor is RUNNING.
- RCP in associated loop -
AVAILABLE AND NOT OPERATING 3) Verify [TV-1IA-400] - OPEN
[1IA-90] Instr Air Vlv To CNMT Instr Air Isol Vlv (West Cable Vault -
735).
- 5) Open all PRZR PORVs and block valves.
(step continued next page) 1FRC1 8/24/2022 13 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 20 (continued from previous page)
- 6) IF core exit TCs remain greater than 1200F, THEN with appropriate keys open reactor coolant vent system valves:
a) Open [SOV-1RC-103A,B],
RC Vent Sys PRZR Vent Isol Vlvs.
b) Open [SOV-1RC-102A,B],
RC Vent Sys Rx Vessel Vent Isol Vlvs.
c) Open [SOV-1RC-105], RC Vent Sys Vent To CNMT Isol Vlv.
- 7) GO TO Step 21.
- d. RETURN TO Step 20.a 21 Depressurize All Intact SGs To Atmospheric Pressure
- a. Dump steam to condenser from a. Manually or locally dump steam intact SG(s) at maximum rate. from intact SGs using:
- SG Atm Dump Valves
-OR-
- IF no ruptured SG exists, THEN use Residual Heat Release Control Valve IF no intact SG available, THEN use faulted or ruptured SG.
Refer to Attachment 2-U, Local Operation of SG Atmospheric Steam Dump Valves.
1FRC1 8/24/2022 14 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 22 Check Five Hottest Core Exit TCs - IF five hottest core exit TCs LESS THAN 1200F dropping, THEN RETURN TO Step 20.
IF five hottest core exit TCs rising AND RCPs running in all available RCS cooling loops, THEN GO TO SAG-1, Initial Response CAUTION IF offsite power is lost after SI reset, manual action may be required to restart safeguards equipment.
23 Check If SI Accumulators Should Be Isolated
- a. LHSI Flow - AT LEAST a. Go to Step 25.
INTERMITTENT FLOW
(step continued next page) 1FRC1 8/24/2022 15 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 23 (continued from previous page)
- c. Insert shorting bars into jacks for [MOV-1SI-865A,B,C]
- d. Close [MOV-1SI-865A,B,C] d. Perform the following:
- 1) Reset CIA and CIB.
- 2) Verify at least one station air compressor or the diesel air compressor is RUNNING.
- 3) Verify [TV-1IA-400] - OPEN
Instr Air Vlv To CNMT Instr Air Isol Vlv (West Cable Vault - 735).
- 5) Vent any unisolated accumulators to atmospheric pressure.
Refer to 1OM-11.4.H, Venting Safety Injection Accumulators
[1SI-TK-1A(1B)(1C)].
IF an accumulator can NOT be isolated or vented, THEN consult the TSC to determine contingency actions.
- e. Remove shorting bars 24 Check If RCPs Should Be Stopped
- a. At least two RCS hot leg a. GO TO Step 25.
temperatures - LESS THAN 350F
- b. Stop all RCPs.
1FRC1 8/24/2022 16 of 17
BVPS - EOP 1OM-53A.1.FR-C.1(ISS3)
Number Title FR-C.1 Response To Inadequate Core Cooling Issue 3 Revision 3 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 25 Verify SI Flow Continue efforts to establish SI flow. Try to establish flow from
-OR-RETURN TO Step 22.
- LSHI Flow - INDICATED 26 Check Core Cooling
- a. At least two RCS hot leg a. RETURN TO Step 22.
temperatures - LESS THAN 350F
- c. RVLIS full range GREATER THAN c. RETURN TO Step 22.
64%.
27 Check RCS Vent Paths - CLOSED
IF any valve can NOT be closed, THEN close its block valve.
- b. Reactor Coolant Vent System b. Manually close valves.
Valves - CLOSED 28 GO TO Step 16 Of E-1, "Loss Of Reactor Or Secondary Coolant"
- END -
1FRC1 8/24/2022 17 of 17
CONTINUOUS ACTION STEPS FR-C.1 STEP DESCRIPTION 1 Check RWST Level - GREATER THAN 19 FEET 9 Check PPDWST Level - GREATER THAN 28.0 FEET 10 Check Intact SG Levels - GREATER THAN 31% [50% ADVERSE CNMT]
11 Check Station Instrument Air Header Pressure - GREATER THAN 100 PSIG
- IF station instrument air header pressure can NOT be restored, THEN check if an AFW Pump should be stopped.
Refer to Attachment 2-S.
Issue 3 Revision 3