Text

Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair
	ML23060A018
Person / Time
Site:	Beaver Valley
Issue date:	03/01/2023
From:	Blair B; Energy Harbor Nuclear Corp
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	L-23-073
	Download: ML23060A018 (1)
	v • d • e

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 1, 2023 L-23-073 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 Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair

Pursuant to 10 CFR 50.90, Energy Harbor Nuclear Corp. hereby requests 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 proposed license amendment is being requested on an emergency basis for BVPS-1 pursuant to 10 CFR 50.91(a)(5). Beaver Valley Power Station, Unit No. 2 (BVPS-2), is not affected by this proposed amendment.

Energy Harbor Nuclear Corp. requests approval of the proposed license amendment as soon as possible and no later than March 3, 2023, to support completion of repair activities.

An evaluation of the proposed change is provided in Attachment 1, including a discussion of the emergency circumstances for BVPS-1. A copy of TS 3.5.2 marked with the proposed change is provided in Attachment 2. A retyped copy of TS 3.5.2 with the proposed change incorporated is provided in Attachment 3 for information only. A copy of the applicable TS Bases pages marked to reflect the proposed change is provided in Attachment 4 for information only. Attachment 5 provides an evaluation of the risk impact related to the proposed license amendment.

Beaver Valley Power Station, Unit No. 1 L-23-073 Page 2

In accordance with Energy Harbor Nuclear Corp. administrative procedures and the quality assurance program manual, this proposed license amendment has been previously reviewed and approved by the plant operation review committee.

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 1, 2023.

Barry N. Blair

Attachments:

1. Evaluation of the Proposed Change

2. Proposed Technical Specification Pages Markup

3. Proposed Technical Specification Pages Re-typed (for information only)

4. Proposed Technical Specification Bases Page Markups (for information only)

5. Risk Analysis for BVPS-1 Emergency TS Change to Address RV-1Sl-857 Leak

cc: NRG Region I Administrator NRG Resident Inspector NRG Project Manager Director BRP/DEP Site BRP/DEP Representative Attachment 1 L-23-073

Evaluation of the Proposed Change Page 1 of 16

Subject:

Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair

1.0

SUMMARY

DESCRIPTION

2.0 DETAILED DESCRIPTION

2.1 System Design and Operation 2.2 Current Technical Specification Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change 2.5 Emergency Circumstances

3.0 TECHNICAL EVALUATION

3.1 Defense-in-Depth 3.2 Safety Margin Evaluation 3.3 Compensatory Measures 3.4 Evaluation of Risk Impacts

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements I Criteria 4.2 No Significant Hazards Consideration Analysis 4.3 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

L-23-073 Page 2 of 16

1.0

SUMMARY

DESCRIPTION

This evaluation supports an emergency license amendment request to amend Renewed Facility Operating License No. DPR-66 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.

In addition, BVPS-1 and Beaver Valley Power Station, Unit No. 2 (BVPS-2) are currently operating under Condition B of TS LCO 3.7.10, Control Room Emergency Ventilation System (CREVS). The repair of the leak would restore the control room envelope (CRE) boundary to operable status and allow TS LCO 3.7.10 to be met for BVPS-1 and BVPS-2 and the associated actions to be exited.

2.0 DETAILED DESCRIPTION

2.1 System Design and Operation

Engineered safety features (ESF), together with the containment system, serve as protection to the public in the unlikely event of a loss-of-coolant accident (LOCA). One of the ESF is the emergency core cooling system (ECCS), which meets the intent of General Design Criteria (GDC) 35 in the 1971 Atomic Energy Commission criteria as noted in the BVPS-1 Updated Final Safety Analysis Report (UFSAR) Section 1.3.2 and Appendix 1A. The ECCS is described in more detail in the BVPS-1 UFSAR Section 6.3.

The ECCS is designed to cool the reactor core, as well as to provide additional shutdown capability following initiation of the following accident conditions:

1. Pipe breaks and spurious relief or safety valve lifting in the reactor coolant system (RCS) which cause a discharge larger than that which can be made up by the normal system, up to and including the instantaneous circumferential rupture of the largest pipe in the RCS.

2. Rupture of a control rod drive mechanism causing a rod cluster control assembly (RCCA) ejection accident.

3. Pipe breaks and spurious relief or safety valve lifting in the steam system, up to and including the instantaneous circumferential rupture of the largest pipe in the steam system.

4. A steam generator tube rupture.

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The primary function of the ECCS for the ruptures described above is to remove the stored and fission product decay heat from the core such that the fuel damage, to the extent that would impair effective cooling of the core, is prevented. This implies that the core remains intact and in place, with its essential heat transfer geometry preserved.

For any rupture of a steam pipe and the associated uncontrolled heat removal from the core, the ECCS will perform such that:

1. In the event of an uncontrolled steam release resulting from any single active failure in the main steam system (such as the opening with failure to close, of any single steam relief, control or bypass valve), there is no return to criticality after reactor trip.

2. In the event of a steam line break together with the combined effects of any single control rod remaining stuck out of the core (after reactor trip) and the most restrictive single active failure in the ESF, the core shall remain in place and intact with its essential heat transfer geometry preserved.

The ECCS is designed to accept a single active failure following the incident without loss of its protective function. The system design will tolerate the failure of any single active component in the ECCS itself or in the necessary service systems at any time during the period of required system operations following the incident.

A single active failure analysis is presented in BVPS-1 UFSAR Table 6.3-1 and demonstrates that the ECCS can sustain the failure of any single active component in either the short or long term and still meet the level of performance for core cooling.

Since the operation of the active components of the ECCS following a steam line rupture is identical to that following a LOCA, the same analysis is applicable and the ECCS can sustain the failure of any single active component and still meet the level of performance for the addition of shutdown reactivity.

For the safety injection system, long-term is the point at which tranfer to recirculation occurs. For long-term emergency core cooling, the system design is based on accepting either a passive or an active failure. The following criteria are utilized in the design of the ECCS:

1. During the long-term cooling period following a LOCA, the emergency core cooling flow paths shall be separable into two subsystems, either of which can provide minimum core cooling functions and return spilled water from the floor of the containment back to the RCS.

2. Either of the two subsystems can be isolated and removed from service in the event of a leak outside the containment.

3. Adequate redundancy of check valves is provided to tolerate failure of a check valve during the long term as a passive component.

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4. Should one of these two subsystems be isolated in this long-term period, the other subsystem remains operable.

5. Provisions are also made in the design to detect leakage from components outside the containment, collect this leakage and to provide for maintenance of the affected equipment.

Thus, for the long-term emergency core cooling function, adequate core cooling capacity exists with an open flow path removed from service whether isolated due to a leak, because of blocking of one flow path, or because failure in the containment results in a spill of the delivery of one subsystem.

The principal components of the ECCS that provide emergency core cooling immediately following a loss of coolant are the three accumulators (one for each loop),

two of the three high head safety injection charging (HHSI) pumps (which perform the charging functions during normal operations), and the two low head safety injection (LHSI) pumps. The HHSI pumps are located in the auxiliary building. The two LHSI pumps are located alongside the containment structure. The accumulators are passive components that discharge into the cold legs of the reactor coolant piping when RCS pressure decreases below accumulator pressure, thus ensuring rapid core cooling for large breaks. The accumulators are located inside the containment.

There are two modes of ECCS operation: (1) injection mode in which any reactivity increase following the postulated accidents is terminated, initial cooling of the core is accomplished and coolant lost from the primary system in the case of a LOCA is replenished; and (2) recirculation mode in which long-term core cooling is provided during the accident recovery period.

The initiation signal for core cooling by the HHSI pumps and the LHSI pumps is the safety injection signal (SIS). The SIS is actuated by any of the following:

1. Low pressurizer pressure (two-out-of-three)

2. High containment pressure (two-out-of-three)

3. Low steam line pressure (two-out-of-three detectors in any one main steam line)

4. Manual actuation (one-out-of-two).

The SIS opens the boron injection header isolation valves and starts the HHSI pumps.

The accumulator isolation valves also receive the SIS, even though these valves are normally open.

The HHSI deliver borated water to the three cold legs of the reactor coolant loops during the injection phase. The suction of the HHSI pumps is diverted from their normal suction at the volume control tank to the refueling water storage tank (RWST) by the SIS. The pumps feed a common injection header. The injection header contains a boron injection tank (BIT) on the discharge side of the HHSI pumps. The HHSI pumps L-23-073 Page 5 of 16

discharge is isolated by the redundant normally closed parallel BIT outlet valves. The valves open upon receipt of an SIS and borated water from the RWST flows from the discharge of the HHSI pumps through the BIT and into the RCS cold legs.

Valves of the safety injection systems that are remotely operated and are normally in their ready position but do not receive a SIS have their positions indicated on a common portion of the main control board that the operator can monitor. At any time during operation when one of these valves is not in the ready position for injection, it is shown visually on the board. In addition, an audible alarm alerts the operator to the condition.

Initial response of the injection systems is automatic, with appropriate allowances for delays in actuation of circuitry and active components. The active portions of the injection systems are automatically actuated by the SIS. In addition, manual actuation of the entire injection system and individual components can be accomplished from the control room. In analysis of system performance, delays in reaching the programmed trip points and in actuation of components were conservatively established on the basis that only emergency onsite power is available. The starting sequence of the HHSI pumps, the LHSI pumps and the related emergency power equipment is designed so that delivery of the full rated flow is reached within 27 seconds after the process parameters reach the setpoints for the injection signal.

The ECCS consists of two redundant, 100 percent capacity trains. Each ECCS train consists of two subsystems: the HHSI subsystem and a LHSI subsystem. The ECCS accumulators, the containment sump, and the RWST are also part of the ECCS, but are not considered part of an ECCS flow path as described by BVPS-1 TS LCO 3.5.2.

The HHSI and LHSI subsystems of each ECCS train are interconnected such that each ECCS train may utilize HHSI or LHSI subsystem components from the other ECCS train. This interconnecting and redundant subsystem design provides the operators with the ability to utilize components from opposite trains to achieve the required 100 percent flow to the core.

For BVPS-1, during the injection phase of a LOCA recovery, a suction header supplies water from the RWST to the ECCS pumps. Water to the HHSI pumps is supplied via parallel motor operated valves to ensure that at least one valve opens on receipt of a safety injection actuation signal. The supply header then branches to the three HHSI pumps. The discharge from the HHSI pumps divides into three supply lines, each of which feeds the injection line to one RCS cold leg. One HHSI pump is dedicated to each train of ECCS. The third pump is a pump that can be substituted for either dedicated HHSI pump in an ECCS train. Throttle valves in the HHSI injection lines are set to balance the flow to the RCS. This balance ensures sufficient flow to the core to meet the analysis assumptions following a LOCA in one of the RCS cold legs.

For LOCAs that are too small to depressurize the RCS below the shutoff head of the LHSI pumps, the HHSI pumps supply water until the RCS pressure decreases below the LHSI pump shutoff head. During this period, the steam generators provide part of the core cooling function.

L-23-073 Page 6 of 16

The HHSI subsystem of the ECCS also functions to supply borated water to the reactor core following increased heat removal events, such as a main steam line break (MSLB).

The HHSI pumps A and B are capable of being automatically started and are powered from separate ESF buses. HHSI pump C can be powered from either of the ESF buses that HHSI pump A or B is powered from. An interlock prevents HHSI pump C from being powered from both ESF buses simultaneously. In the event of a safety injection actuation signal coincident with a loss of offsite power, interlocks prevent operation of two HHSI pumps on the same bus to prevent overloading the EDGs.

The transfer from the safety injection mode to the recirculation mode will automatically take place on an extreme low-level signal from the RWST. At the termination of the injection phase, the suction of the LHSI pumps is realigned from the RWST to the containment sump, the LHSI pumps minimum recirculation valves are closed, and the valves in the charging pumps suction are realigned from the RWST to the discharge of the LHSI pumps.

2.2 Current Technical Specification Requirements

LCO 3.5.2 requires that two ECCS trains shall be operable. This LCO is applicable in Modes 1, 2, and 3. Note 3, which is for BVPS-1 only, states that in Mode 3 the ECCS HHSI flow path may be isolated to support transition into or from the Applicability of LCO 3.4.12, Overpressure Protection System (OPPS) for up to four hours or until the temperature of all RCS cold legs exceeds the OPPS enable temperature specified in the Pressure and Temperature Limits Report (PTLR) plus 25°F, whichever comes first.

With the plant operating in Mode 1, Condition A is entered if one or more trains of ECCS are inoperable. The train(s) must then be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or Condition B is entered, which requires the plant to be in Mode 3 within six hours and Mode 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If there is less than 100 percent of the ECCS flow equivalent to a single operable ECCS train, the plant would be in Condition C. Condition C requires immediate entry into LCO 3.0.3.

Surveillance Requirement (SR) 3.5.2.3 requires verification that each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.2.5 requires verification that each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

L-23-073 Page 7 of 16

2.3 Reason for the Proposed Change

On February 12, 2023, a leak was discovered that appeared to be ECCS water from BIT relief valve RV-1SI-857. Sampling for boron concentration supported that the leak was from the ECCS. The leakage exceeded the stated surveillance limit for the radiological dose to the control room, which is common to BVPS-1 and BVPS-2. As a result, both BVPS-1 and BVPS-2 entered TS LCO 3.7.10 on February 12, 2023.

Immediate actions were implemented with both units isolating the control room dampers and monitoring the leak degradation daily. As a mitigating action, an engineering evaluation report (EER) was approved, which provides a revised ECCS leakage limit of 45,600 cubic centimeter per hour (cc/hr). This additional analysis credits Supplementary Leak Collection and Release System (SLCRS) being in operation. This revised ECCS leakage limit satisfies the 24-hour action for LCO 3.7.10 Condition B, Required Action B.2, but does not support control room envelope (CRE) operability. As such, the 90-day completion time for LCO 3.7.10 Required Action B.3 to restore the CRE boundary to Operable status still applies and will expire on May 13, 2023, at 0800.

While mitigating actions are in place until the leak can be repaired, the leakage has continued to increase toward the current ECCS allowable leakage value. The emergency circumstances involved with this proposed license amendment are provided in Section 2.5 of this request.

While margin currently exists in the amount of ECCS leakage, leakage continues to increase and is expected to continue to increase. Operating experience with similar relief valves under similar circumstances in the industry showed continued degradation of the leak rate. A probability risk assessment (PRA) analysis was performed, and the results show that there is low risk associated with performing the needed repair while the plant is in Mode 1. This approach allows for resolution of the leak, allows TS LCO 3.7.10 to be met, and avoids an unnecessary shutdown of BVPS-1 without a corresponding public health and safety benefit.

2.4 Description of the Proposed Change

The normal safety injection (SI) flow path passes through manual valves 1SI-867A, BIT Inlet Isolation Valve, and 1SI-867B, BIT Inlet Isolation Valve (in parallel), then through the BIT, and through the automatically opening MOV-1SI-867C, BIT Outlet Isolation Valve, and MOV-1SI-867D, BIT Outlet Isolation Valve (in parallel), following an SIS into the RCS cold legs. Isolating and repairing the leaking RV-1SI-857, BIT Relief Valve, while in Mode 1 would require closing valves 1SI-867A, 1SI-867B, MOV-1SI-867C, and MOV-1SI-867D, thereby isolating the normal flow path.

An alternate flow path is available through MOV-1SI-836 (that bypasses the BIT) into the RCS cold legs. However, this valve does not automatically open and would require a manual operator action from the control room bench board to open following a SIS.

To repair the ECCS leakage from BIT relief valve RV-1SI-857 in Mode 1, the alternate manual flow path is needed to bypass the BIT.

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The alternate manual flow path through MOV-1SI-836 is not considered redundant.

Although the flow path would enable the repair to be made to the BIT, if MOV-1SI-836 failed shut, it would result in a loss of the safety injection function because the flow path is not single failure proof. Thus, a train (Train B) of ECCS would be declared inoperable.

The duration of the repair is expected to be no more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The PRA analysis shows the duration of the low-risk as 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Therefore, the proposed change includes a restriction not to exceed 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . With compensatory actions in place during the repair activity, it is proposed that the alternate manual flow path be allowed for use during this time. The following note is proposed to be added to LCO 3.5.2:

4. For Unit 1 only. One ECCS train may use an alternate manual flow path on a one-time basis not to exceed 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months while the compensatory measures described in Section 3.3 of Energy Harbor Nuclear Corp. letter L-23-073, dated March 1, 2023, are implemented, if not otherwise inoperable. This allowance expires at 2400 EDT on April 7, 2023.

With the addition of Note 4, some content on the TS pages will shift to the following page. However, no changes are made to that content.

2.5 Emergency Circumstances

Why the Condition Occurred:

On February 12, 2023, at 0800, the BVPS-1 and BVPS-2 control room emergency ventilation systems (CREVS) were declared inoperable due to the safety injection relief valve (discharging or leaking) to the primary auxiliary building. This leakage, in conjunction with a design-basis loss-of-coolant accident, may result in radiological dose exceeding limits to the exclusion area boundary and to the control room, which is common to both BVPS-1 and BVPS-2. The events leading to these emergency circumstances are documented below.

On February 12, 2023, at 0300, a Radiation Protection Technician reported a leak coming from the BIT cubicle.

On February 12, 2023, at 0515, it was documented that the leak appeared to be ECCS water coming from a cracked weld cap on the tailpipe downstream of RV-1SI-857. The leakage was sampled by Chemistry and the associated boron concentration supported that the leakage was ECCS water. Strategic Engineering was contacted to quantify the leakage.

On February 12, 2023, at 0800, the leak coming from RV-1SI-857 was quantified.

The leakage values noted equated to 16,800 cc/hr, which exceeds the bounding UFSAR ECCS leakage limit of 11,400 cc/hr (UFSAR Section 14.3.5.2).

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On February 14, 2023, a temporary modification was approved to install a relief valve (RV-1SI-857A) to provide system over pressurization relief in lieu of RV-1SI-857.

On February 16, 2023, the alternate relief valve RV-1SI-857A was installed. In an attempt to eliminate or reduce leakage, RV-1SI-857 was gagged finger tight plus one quarter turn (determined later to be 4 ft-lbs). Initially, the leakage was slightly reduced as a result of the gag. Subsequently, the leak rate began to increase, and additional torque was applied on February 23 and 24, 2023.

Torque was increased in 5 ft-lb increments, while monitoring leakage, up to the recommended torque on the RV-1SI-857 gag of 19 ft-lbs in accordance with an EER. Since torquing to 19 ft-lbs, the leakage has continued to degrade. This degradation is likely due to continued leakage cutting of the relief valve seat.

On February 16, 2023, an EER was approved for updated ECCS allowable leakage value while crediting SLCRS. The new ESF leakage acceptance criteria of 45,600 cc/hr assumes two HHSI pumps are running at design-basis accident pressure. Crediting the SLCRS ensures that the control room operator dose consequences will remain below the current design analysis value and offsite doses would remain at their current values.

Both BVPS-1 and BVPS-2 are in TS LCO 3.7.10, Condition B. Required Action B.3 is to restore the CRE boundary to Operable status within a 90-day Completion Time that expires on May 13, 2023, at 0800. Energy Harbor Nuclear Corp. has been investigating different options on how to stop the leakage from RV-1SI-857. In addition to the gag of RV-1SI-857 discussed above, the following options have been considered:

Freeze Seal as Single-Point-Clearance - The freeze seal alone without isolation will not be used due to personnel and nuclear safety concerns. The ECCS is designed to automatically actuate causing changes to system pump and valve alignments. The system pressure and temperatures are not constant during different accident conditions and with the high pressures of the system it would not be prudent to accept the risk involved in this evolution.

In addition, the applicable site procedure lists 120 pounds per square inch differential (psid) as the maximum acceptable pressure without further evaluation. The maximum pipe surface temperature is listed as 120°F.

Referencing Electrical Power Research Institute (EPRI) Technical Report TR-016384-R1, general guidelines for system and differential pressure across the ice plug are less than 400 pounds per square inch gauge (psig) and fluid temperatures of less than 130°F. An industry search was performed; no operational experience (OE) was found that was relevant to 2575 psig system pressure and temperatures, which could exceed 130°F in a design-basis accident. The evaluation for the freeze seal to act as the Class 2 piping pressure boundary with the system in-service during normal and accident conditions would have been extensive and may have been found not to be acceptable.

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Line Stop - Line Stopping is a means of temporarily stopping flow in an operating pipe and can be used to isolate piping systems for repairs, alteration or relocations without shutdown or loss of service. If used in conjunction with bypass lines, product flow can be continued around the isolated section of the one-way feed pipes under repair. This is not a viable option due to spacing restraints and the potential for foreign material intrusion into the ECCS. As such, this option was eliminated.

Adding the SLCRS back into the Licensing Basis - The SLCRS has not been credited in the current dose analysis since 2003 when BVPS implemented the alternate source term methodology. To add SLCRS back into the licensing basis, the dose analyses would need to be re-performed by Westinghouse, followed by license amendment request development, Nuclear Regulatory Commission (NRC) staff review and approval. In addition, with the possibility of continued leak degradation, this would not be a long-term success path.

Follow-up Operability Determination to restore the CRE to Operable - This is not a viable long-term option considering continued leak degradation.

Shutdown Unit 1 to Mode 4 - The PRA analysis indicates that the risk of crediting the alternate ECCS flow path will be minimal to perform the work in Mode 1.

Granting of the requested amendment would avoid a potentially unnecessary plant transient and shutdown of a reactor without a corresponding health and safety benefit. Therefore, performing the work in Mode 1 is the preferred path to address the leakage.

Next Steps:

As of February 19, 2023, daily leak rate monitoring has been conducted in accordance with the site procedure to collect RCS pressure boundary leakage and manual collections, which are to quantify leakage specific to RV-1SI-857. Even with the additional torques on the gag that were performed on February 23 and 24, 2023, daily leakage values have continued to increase toward the updated ECCS allowable leakage value. If this value is exceeded, it would require both BVPS-1 and BVPS-2 to enter LCO 3.7.10 Condition C requiring shutdown to Mode 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This would prompt isolation of the BVPS-1 BIT. Isolation of the BIT would require BVPS-1 to enter LCO 3.5.2 Condition C, which requires immediate entry into LCO 3.0.3.

If the leakage continues to degrade as expected, plant conditions would not support the timelines required by either a routine or an exigent license amendment. Without an emergency amendment, BVPS-1 would shutdown to Mode 4 to install a blank flange on RV-1SI-857.

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Blank Flange Installation Based on NRC Approval of Emergency LAR:

RV-1SI-857 will be removed from the system and a blank flange(s) spool pieces will be put on the exposed piping ends.

A freeze seal will be used for water management during this installation.

Why the Situation Could Not Be Avoided:

The causal analysis for the leak is still in progress to determine and leverage additional learnings. Early indications are that the tailpipe on RV-1SI-857 may have become clogged and the cap cracked due to increased pressure.

RV-1SI-857 has been maintained within its normal preventative maintenance frequency and showed no signs of leakage when exiting the fall refueling outage in November 2022. The first indication of leakage was in February 2023.

3.0 TECHNICAL EVALUATION

3.1 Defense-in-Depth

During the time that LCO 3.5.2 Note 4 is invoked, the alternate safety injection flow path will remain operable and available. In the event that an automatic safety injection occurs, the manual valve (MOV-1SI-836) will be opened by a dedicated Reactor Operator. Should an event occur that requires a manual safety injection, safety injection will be actuated by the Reactor Operator (At The Controls) and MOV-1SI-836 will be manually opened by the extra Reactor Operator dedicated to this task.

Additionally, an extra, dedicated operator will be assigned to manually open MOV-1SI-836 locally (in the event that MOV-1SI-836 fails to stroke from the control room). This operator will be stationed locally in 722 Safeguards Penetrations area and will have received a High Radiation Area briefing to perform the applicable task prior to invoking LCO 3.5.2 Note 4. In addition, compensatory measures as described below will be in place and available.

3.2 Safety Margin Evaluation

The proposed TS change is consistent with the principle that sufficient safety margins are maintained based on the defense-in-depth described in Section 3.1 and the compensatory measures described in Section 3.3.

3.3 Compensatory Measures

The following compensatory measures are required during the time period that LCO 3.5.2 Note 4 is invoked.

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Normal System Arrangement:

1. MOV-1SI-867C and MOV-1SI-867D are energized closed motor operated valves.

Both valves will automatically open on a Safety Injection (SI) Signal. Both valves are also capable of being opened manually via control switch on control room bench board.

2. 1SI-867A and 1SI-867B are Locked Open manual valves.

3. MOV-1SI-836 is an energized closed motor operated valve. The valve does not receive a safety injection signal. The valve is capable of being opened manually via control switch on control room bench board.

Alternate SI Alignment with LCO 3.5.2 Note 4 Invoked:

1. MOV-1SI-867C and MOV-1SI-867D will both be closed. The 480V power supply for both motor-operated valves will be deenergized. Both valves will serve as a clearance boundary for repair of RV-1SI-857.

2. 1SI-867A and 1SI-867B will both be closed. Both valves will serve as a clearance boundary for repair of RV-1SI-857.

3. 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.

Additional compensatory measures to maintain a defense-in-depth posture will be:

- No redundant Technical Specification equipment will be removed from service.

The spare C river water pump is out of service to support a planned maintenance activity. This activity is considered in the PRA analysis and does not represent a significant increase in configuration risk. Both trains of Engineered Safety Features (ESF) equipment will be operable and available with the exception of the boron injection flow path outage to support this repair.

- No generation risk activities will be planned or authorized.

- 1-CR-4, Process Instrumentation Room, and 1-ES-1, Train A Emergency Switchgear Room, will have continuous fire watches and no hot work will be permitted.

- Work on the 345kV and 138kV will be restricted and no switching activities will be permitted.

- Control room operators will be briefed on the importance of beginning to cooldown and depressurize in accordance with EOP Network if MOV-1SI-836 should fail to open.

L-23-073 Page 13 of 16

- Dedicated Reactor Operator to open MOV-1SI-836 in the control room if a safety injection is required.

- Dedicated field operator located in safeguards to manually open MOV-1SI-836 if required.

3.4 Evaluation of Risk Impacts

The risks associated with this one-time action have been evaluated by way of probabilistic risk assessment (PRA) Effective Reference Model PRA-BV1-AL-R08. This is the PRA model of record issued on January 5, 2023.

This plant-specific risk assessment followed the guidance in Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, and RG 1.177, An Approach for Plant-Specific, Risk-Informed Decision Making: Technical Specifications. of this submittal presents the evaluation of risk impacts due to the proposed license amendment.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements / Criteria

All features of BVPS-1 related to public safety are designed to the single failure criterion. One of the fluid systems to which the single failure criterion applies, as listed in Appendix A of 10 CFR 50 is Emergency Core Cooling. BVPS-1 was designed and constructed to comply with 1967 Atomic Energy Commission (AEC) General Design Criteria (GDC) as noted in the BVPS-1 Updated Final Safety Analysis Report (UFSAR)

Section 1.3.2. BVPS-1 UFSAR Appendix 1A provides a discussion of the degree of conformance to the AEC GDC published as Appendix A to 10 CFR 50 in July 1971.

The numbering of the GDCs differs between the two appendices. For clarity, the 1971 number for Emergency Core Cooling (GDC 35) is provided.

Emergency Core Cooling (GDC 35)

Criterion

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 L-23-073 Page 14 of 16

system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Design Conformance

The BVPS-1 design conforms with the intent of criterion 35. Appropriate core cooling systems have been designed so as to provide for the removal of core thermal loads and for the limiting of metal water reactions to an insignificant level. Suitable redundancy is provided in core cooling systems. The charging/safety injection, accumulator and safety injection systems will accommodate a single active failure and still fulfill their intended safety function.

10 CFR 36, Technical specifications

Regulatory requirements related to the contents of the TS. Specifically, 10 CFR 50.36(c)(2) states, in part, Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

The ECCS trains (HHSI subsystem) satisfy 10 CFR 50.36(c)(2)(ii), Criterion 3, which states the following:

(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The ECCS, including the HHSI subsystem, is described in the BVPS-1 UFSAR Section 6.3.

The proposed license amendment does not delete requirements associated with the system and LCO 3.5.2 continues to maintain requirements associated with structures, systems, and components that are part of the primary success path and actuate to mitigate the related design-basis accidents and transients. The proposed amendment does not alter the remedial actions or shutdown requirements required by 10 CFR 50.36(c)(2)(i). The proposed change does not affect compliance with this regulation.

Following implementation of the proposed change, BVPS-1 will remain in compliance with applicable design criteria as described in the BVPS-1 UFSAR.

L-23-073 Page 15 of 16

4.2 No Significant Hazards Consideration Analysis

Energy Harbor Nuclear Corp. is proposing to amend the Renewed Facility Operating License No. DPR-66 for Beaver Valley Power Station, Unit No. 1 (BVPS-1). The amendment 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 valve. 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 valve.

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

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

Response: No.

The proposed change allows a one-time use of an alternate manual flow path in the Emergency Core Cooling System (ECCS). The proposed change does not affect accident initiators or precursors. The ECCS will remain capable of adequately responding to a design basis event or transient during the period that the note is invoked. A probability risk assessment (PRA) was performed for this proposed change and determined that this has low risk.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The function of ECCS is to cool the core as well as provide additional shutdown capability following an accident. ECCS operation is not a precursor for any accident listed in Chapter 14 of the Updated Final Safety Analysis Report. The proposed change allows a one-time use of an alternate manual flow path in the ECCS. The function of the ECCS is maintained during the period that the note is invoked.

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

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

Response: No.

L-23-073 Page 16 of 16

The proposed change allows a one-time use of an alternate manual flow path ECCS.

The proposed change does not exceed or alter a design basis or safety limit. 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. The alternate injection flow path is validated in the High Head Safety Injection Full Flow test ensuring the proper injection flowrate. The alternate flow path is the same safety class as the normal flow path and the alternate injection valve receives emergency power.

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

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

4.3 Conclusions

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

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Attachment 2 L-23-073

Proposed Technical Specification Pages Markup (5 pages follow)

ECCS - Operating 3.5.2

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

3.5.2 ECCS - Operating

LCO 3.5.2 Two ECCS trains shall be OPERABLE.

- NOTES -

1. In MODE 3, both low head safety injection pump flow paths may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

2. In MODE 3, one of the required charging pumps may be made incapable of injecting to support transition into or from the Applicability of LCO 3.4.12, "Overpressure Protection System (OPPS)," for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or until the temperature of all RCS cold legs exceeds the OPPS enable temperature specified in the PTLR plus 25F, whichever comes first.

3. For Unit 1 only. In MODE 3, the ECCS automatic high head safety injection (HHSI) flow path may be isolated to support transition into or from the Applicability of LCO 3.4.12, "Overpressure Protection System (OPPS)" for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or until the temperature of all RCS cold legs exceeds the OPPS enable temperature specified in the PTLR plus 25F, whichever comes first.

4. For Unit 1 only. One ECCS train may use an alternate manual flow path on a one-time basis not to exceed 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months while the compensatory measures described in Section 3.3 of Energy Harbor Nuclear Corp. letter L-23-073, dated March 1, 2023, are implemented, if not otherwise inoperable. This allowance expires at 2400 EDT on April 7, 2023.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. One or more trains A.1 Restore train(s) to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND

B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE ECCS train available.

Beaver Valley Units 1 and 2 3.5.2 - 1 Amendments 278 / 161 ECCS - Operating 3.5.2

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND

B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE ECCS train available.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

SR 3.5.2.1 Verify the following valves are in the listed position with In accordance power to the valve operator control circuit removed. with the Surveillance Frequency Control Program

For Unit 1 only

Number Position Function MOV-1SI-890A Closed Low head safety injection (LHSI) to Hot Leg MOV-1SI-890B Closed LHSI to Hot Leg MOV-1SI-890C Open LHSI to Cold Leg MOV-1SI-869A Closed HHSI Pump to Hot Leg MOV-1SI-869B Closed HHSI Pump to Hot Leg

For Unit 2 only

Number Position Function 2SIS*MOV8889 Closed LHSI to Hot Legs 2SIS*MOV869A Closed HHSI to Hot Leg 2SIS*MOV869B Closed HHSI to Hot Leg 2SIS*MOV841 Open HHSI to Cold Leg 2CHS*MOV8132A Open HHSI Pump Discharge Cross Connect 2CHS*MOV8132B Open HHSI Pump Discharge Cross Connect 2CHS*MOV8133A Open HHSI Pump Discharge Cross Connect 2CHS*MOV8133B Open HHSI Pump Discharge Cross Connect SR 3.5.2.2 Verify the HHSI pump minimum flow valve is open with In accordance power to the valve operator removed. with the Surveillance Frequency Control Program

SR 3.5.2.3 Verify each ECCS manual, power operated, and In accordance automatic valve in the flow path, that is not locked, with the sealed, or otherwise secured in position, is in the Surveillance correct position. Frequency Control Program

Beaver Valley Units 1 and 2 3.5.2 - 2 Amendments 292 / 179 ECCS - Operating 3.5.2

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

SR 3.5.2.2 Verify the HHSI pump minimum flow valve is open with In accordance

power to the valve operator removed. with the

Surveillance

Frequency

Control Program

SR 3.5.2.3 Verify each ECCS manual, power operated, and In accordance

automatic valve in the flow path, that is not locked, with the

sealed, or otherwise secured in position, is in the Surveillance

correct position. Frequency

Control Program

SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance

flow point is greater than or equal to the required with the

developed head. INSERVICE

TESTING

PROGRAM

SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that In accordance

is not locked, sealed, or otherwise secured in position, with the

actuates to the correct position on an actual or Surveillance

simulated actuation signal. Frequency

Control Program

SR 3.5.2.6 Verify each ECCS pump starts automatically on an In accordance

actual or simulated actuation signal. with the

Surveillance

Frequency

Control Program

Beaver Valley Units 1 and 2 3.5.2 - 3 Amendments 307 / 197 Attachment 3 L-23-073

Proposed Technical Specification Pages Re-typed (for information only)

(3 pages follow)

ECCS - Operating 3.5.2

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

3.5.2 ECCS - Operating

LCO 3.5.2 Two ECCS trains shall be OPERABLE.

- NOTES -

1. In MODE 3, both low head safety injection pump flow paths may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

2. In MODE 3, one of the required charging pumps may be made incapable of injecting to support transition into or from the Applicability of LCO 3.4.12, "Overpressure Protection System (OPPS)," for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or until the temperature of all RCS cold legs exceeds the OPPS enable temperature specified in the PTLR plus 25F, whichever comes first.

3. For Unit 1 only. In MODE 3, the ECCS automatic high head safety injection (HHSI) flow path may be isolated to support transition into or from the Applicability of LCO 3.4.12, "Overpressure Protection System (OPPS)" for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or until the temperature of all RCS cold legs exceeds the OPPS enable temperature specified in the PTLR plus 25F, whichever comes first.

4. For Unit 1 only. One ECCS train may use an alternate manual flow path on a one-time basis not to exceed 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months while the compensatory measures described in Section 3.3 of Energy Harbor Nuclear Corp. letter L-23-073, dated March 1, 2023, are implemented, if not otherwise inoperable. This allowance expires at 2400 EDT on April 7, 2023.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. One or more trains A.1 Restore train(s) to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

Beaver Valley Units 1 and 2 3.5.2 - 1 Amendments 278 / 161 ECCS - Operating 3.5.2

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND

B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE ECCS train available.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

SR 3.5.2.1 Verify the following valves are in the listed position with In accordance power to the valve operator control circuit removed. with the Surveillance Frequency Control Program

For Unit 1 only

Number Position Function MOV-1SI-890A Closed Low head safety injection (LHSI) to Hot Leg MOV-1SI-890B Closed LHSI to Hot Leg MOV-1SI-890C Open LHSI to Cold Leg MOV-1SI-869A Closed HHSI Pump to Hot Leg MOV-1SI-869B Closed HHSI Pump to Hot Leg

For Unit 2 only

Number Position Function 2SIS*MOV8889 Closed LHSI to Hot Legs 2SIS*MOV869A Closed HHSI to Hot Leg 2SIS*MOV869B Closed HHSI to Hot Leg 2SIS*MOV841 Open HHSI to Cold Leg 2CHS*MOV8132A Open HHSI Pump Discharge Cross Connect 2CHS*MOV8132B Open HHSI Pump Discharge Cross Connect 2CHS*MOV8133A Open HHSI Pump Discharge Cross Connect 2CHS*MOV8133B Open HHSI Pump Discharge Cross Connect

Beaver Valley Units 1 and 2 3.5.2 - 2 Amendments 292 / 179 ECCS - Operating 3.5.2

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

SR 3.5.2.2 Verify the HHSI pump minimum flow valve is open with In accordance

power to the valve operator removed. with the

Surveillance

Frequency

Control Program

SR 3.5.2.3 Verify each ECCS manual, power operated, and In accordance

automatic valve in the flow path, that is not locked, with the

sealed, or otherwise secured in position, is in the Surveillance

correct position. Frequency

Control Program

SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance

flow point is greater than or equal to the required with the

developed head. INSERVICE

TESTING

PROGRAM

SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that In accordance

is not locked, sealed, or otherwise secured in position, with the

actuates to the correct position on an actual or Surveillance

simulated actuation signal. Frequency

Control Program

SR 3.5.2.6 Verify each ECCS pump starts automatically on an In accordance

actual or simulated actuation signal. with the

Surveillance

Frequency

Control Program

Beaver Valley Units 1 and 2 3.5.2 - 3 Amendments 307 / 197 Attachment 4 L-23-073

Proposed Technical Specification Bases Page Markups (for information only)

(2 pages follow)

ECCS - Operating B 3.5.2

BASES

LCO (continued)

During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the ECCS pumps and their respective supply headers to each of the three cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to supply its flow simultaneously to both the RCS hot or cold legs for Unit 1. The flow path from the containment sump is cycled alternatively between the RCS cold legs or hot legs for Unit 2.

The flow path for each train must maintain its designed independence to ensure that no single failure can disable both ECCS trains.

The LCO is modified by threefour Notes. Note 1 provides an exception allowing the LHSI flow paths to be isolated for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in MODE 3, under controlled conditions, to perform pressure isolation valve testing per SR 3.4.14.1. The flow path is readily restorable from the control room.

As indicated in Note 2, operation in MODE 3 with one required charging pump made incapable of injecting in order to facilitate entry into or exit from the Applicability of LCO 3.4.12, "Overpressure Protection System (OPPS)," is necessary when OPPS enable temperature is at or near the MODE 3 boundary temperature of 350F. LCO 3.4.12 requires that one required charging pump be rendered incapable of injecting at and below the OPPS enable temperature. When this temperature is at or near the MODE 3 boundary temperature, time is needed to make a required charging pump incapable of injecting prior to entering the OPPS Applicability, and provide time to restore the inoperable pump to OPERABLE status on exiting the OPPS Applicability.

Note 3 is only applicable to Unit 1. As indicated in Note 3, operation in MODE 3 with the Unit 1 ECCS automatic high head safety injection (HHSI) flow path isolated in order to facilitate entry into or exit from the Applicability of LCO 3.4.12, "Overpressure Protection System (OPPS)," is necessary when the OPPS enable temperature is at or near the MODE 3 boundary temperature of 350F. LCO 3.4.12 requires that the Unit 1 ECCS automatic HHSI flow path be isolated when any RCS cold leg temperature is the enable temperature specified in the PTLR. When this temperature is near the MODE 3 boundary temperature, Note 3 provides time to isolate the ECCS automatic HHSI flow path prior to entering the OPPS Applicability, and to restore the flow path on exiting the OPPS Applicability.

Note 4 is only applicable to Unit 1. As indicated in Note 4, one ECCS train may use an alternate manual flow path on a one-time basis not to exceed 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months while the compensatory measures described in Section 3.3 of Energy Harbor Nuclear Corp. letter L-23-073, dated March 1, 2023, are implemented, if not otherwise inoperable. This allowance expires in April 2023.

Beaver Valley Units 1 and 2 B 3.5.2 - 6 Revision 0 Attachment 5 L-23-073

Risk Analysis for BVPS-1 Emergency TS Change to Address RV-1SI-857 Leak (24 pages follow)

PRA-BV1-23-003-R0O NOBP-CC-6002, Rev. 05, Alt. 5 Page 1 of24 PRA APPLICATIONS ANALYSIS/ASSESSMENT COVERSHEET

Analysis/Assessment Sequence No.: PRA-BV1-23-003-R00 Rev.:.QQ__

Ref. PRA Tracking#: _ __,_N=/A ________________ (if applicable)

Subject:

Risk Analysis for BVPS-1 Emergency TS Change to Address RV-1Sl-857 Leak

==

Description:==

This PRA Applications Analysis/Assessment provides the PRA risk basis for the one-time, emergency Technical Specification 3.5.2 change request to the NRC to support the permanent removal of the leaking safety injection (SI) relief valve RV-1SI-857. The relief valve removal will require the normal SI alignment to be out-of-service, and the alternate SI alignment to be manually available given the need for SI injection. Quantitative and qualitative risk discussions are provided.

Documents Used by this Analysis/Assessment:

Beaver Valley Technical Specification 3.5.2, Regulatory Guide 1.177 Rev 2, Condition Report CR-2023-00971. NORM-LP-4002 Rev 3

Documents Supported by this Analysis/Assessment:

LAR Submittal L-23-073

Documents Superseded by this Analysis/Assessment:

None

Preparer: Richard J. Stremple -~ fJit2 Date:,;;./~g /;).3 Reviewer: Daniel R. Weller ~ ~,* ~A Date: <- / lJ Id.]

Additional Reviews (If required) Performed by:

Date:

Approved: Robert Drsek for K. Raymond Fine ~~d'Ji:~/vbate: '1__./l&/23 Supervisor Analytical Methods (PRA) or Designee PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 2 of 24

PRA APPLICATIONS ANALYSIS/ASSESSMENT TABLE OF CONTENTS

1.0 Purpose....................................................................................................................... 3 2.0 Description.................................................................................................................. 3 3.0 Guidelines................................................................................................................... 3 4.0 Methodology................................................................................................................ 4 5.0 Assumptions................................................................................................................ 4 6.0 Analysis....................................................................................................................... 5 Model Basis................................................................................................................. 5 Modeling...................................................................................................................... 5 Top Risk Contributors.................................................................................................. 6 Sensitivity Cases.......................................................................................................... 8 Sensitivity Case 1........................................................................................................ 8 Sensitivity Case 2........................................................................................................ 9 Sensitivity Case 3...................................................................................................... 10 7.0 ASME PRA Standard Evaluation............................................................................... 11 8.0 Results...................................................................................................................... 11

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 3 of 24 1.0 Purpose

The purpose of this analysis is to provide the PRA risk basis input into the one-time, emergency Technical Specification 3.5.2 (ECCS - Operating) change request to the NRC to support the permanent removal of leaking safety injection (SI) relief valve RV-1SI-857.

The relief valve removal will require the normal SI alignment to be out-of-service, and the alternate SI alignment to be manually available given the need for SI injection.

2.0 Description

Safety injection (SI) relief valve RV-1SI-857 is leaking within the normal SI makeup and attempts to gag and torque the valve to prevent leakage have not been successful (Condition Report CR-2023-00971). ECCS leakage poses a long-term challenge to control room habitability during accident conditions. There are no simplistic means of isolating the relief valve from the high-pressure charging system, and freeze sealing at operating temperatures and pressures poses potential safety issues to personnel. The decision was made to completely remove RV-1SI-857 before conditions significantly degrade, which will require the normal SI alignment to be out-of-service, and the alternate SI alignment to be manually available given the need for SI injection. For this to occur, a one-time, emergency Technical Specification 3.5.2 (ECCS - Operating) change request will be submitted to the NRC to support the permanent removal of leaking safety injection (SI) relief valve RV-1SI-857.

3.0 Guidelines

NRC Regulatory Guide 1.174 provides an approach that is acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for developing risk-informed applications for a licensing basis change that considers engineering issues and applies risk insights. It provides general guidance concerning analysis of the risk associated with proposed changes in plant design and operation.

NRC Regulatory Guide 1.177 provides an approach that is acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for developing risk-informed applications for changes to completion times (CTs) and surveillance frequencies (SFs) of plant technical specifications (TS). This RG provides specific guidance for considering engineering issues and using risk information to evaluate nuclear power plant TS changes to CTs and SFs.

Section 2.4 (Acceptance Guidelines for Technical Specification Changes) of Regulatory Guide 1.177 Rev. 2 provides (in part) the following related to one-time CT changes:

The following TS acceptance guidelines specific to one-time only CT changes are provided for evaluating the risk associated with the revised CT:

a. The licensee has demonstrated that implementation of the one-time only TS CT change impact on plant risk is acceptable (Tier 1):

1 ICCDP of less than 1.0x10-6 and an ICLERP of less than 1.0x10-7, or PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 4 of 24 2 ICCDP of less than 1.0x10-5 and an ICLERP of less than 1.0x10-6 with effective compensatory measures implemented to reduce the sources of increased risk.

b. The licensee has demonstrated that there are appropriate restrictions on dominant risk-significant configurations associated with the change (Tier 2).

c. The licensee has implemented a risk-informed plant configuration control program, including the procedures to use, maintain, and control such a program (Tier 3).

PRA technical adequacy is addressed through NRC Regulatory Guide 1.200, which references the ASME PRA standard, RA-Sa-2009.

4.0 Methodology

This assessment uses the BV1REV8 PRA model to evaluate the risk associated with the proposed repair configuration invoking the one-time TS CT change against the acceptance criteria outlined in Regulatory Guide 1.177.

The Effective Reference Model PRA-BV1-AL-R08 (BV1REV8) is used as the starting point, from which specific application models are cloned to model the proposed plant configuration for this evolution. The delta CDF and delta LERF are calculated for this configuration, from which the ICCDP and ICLERP are subsequently calculated to determine the acceptability of the proposed one-time CT change in accordance with the guidelines in Regulatory Guide 1.177. Additional application models are created for sensitivity cases, to evaluate the impact of uncertainties on the results. Specific changes made to model the proposed configuration and the sensitivity cases are provided in Section 6.0 of this assessment.

5.0 Assumptions

5.1 The proposed repair configuration will close manual valves 1SI-867A and 1SI-867B, and will de-energize the MCC power supplies to valves MOV-1SI-867C and MOV-1SI-867D, in order to isolate flow to RV-1SI-857 so that the work may be performed without undue personal safety risk.

5.2 Dedicated operators will be stationed both at the benchboard control for MOV-1SI-836 and at MOV-1SI-836 in the field, in order to ensure that the valve can be opened if Safety Injection flow is required during the repair evolution. Despite this, the risk assessment will conservatively use the nominal HEPs for the operator actions to open the valve. Modified HEPs will only be credited in sensitivity cases.

5.3 At the time this assessment is prepared, the station intends to ask for a one-time change to Technical Specification 3.5.2 with a completion time (CT) of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . While this assessment determines the maximum CT that could be supported by the PRA in accordance with the ICCDP and ICLERP acceptance criteria specified in Regulatory Guide 1.177, the ICCDP and ICLERP values calculated for this intended request of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months are also provided.

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 5 of 24 5.4 The only additional planned maintenance on PRA components to occur while the plant is in this proposed configuration will be cleaning of Intake Bay C, making River Water pump 1WR-P-1C unavailable. No generation risk activities will be planned or authorized.

6.0 Analysis

Model Basis

PRA Effective Reference Model: PRA-BV1-AL-R08 (BV1REV8) - This is the PRA average maintenance model of record issued on January 5, 2023 and is used as the basis for this assessment.

RISKMAN For Windows version 15.3 was used for quantification and reports.

Modeling (RG 1.177 Tier 1)

The BVPS-1 Effective Reference PRA model BV1REV8 (PRA -BV1-AL-R08) was cloned and renamed as B1R8HCL1. The following basic event changes were made in B1R8HCL1 to model the proposed repair configuration:

HVXCSI867A = 1.0

HVXCSI867B = 1.0

These are the transfer closed failure mode basic events for manual valves 1SI-867A and 1SI-867B in the normal Safety Injection path. Setting these basic events to 1.0 fails flow through the normal safety injection flowpath in the PRA model, leaving only the alternate flowpath through MOV-1SI-836 as the sole means of providing high pressure injection to the core. MOV-1SI-836 does not open automatically on a Safety Injection signal and must be manually opened by an operator. This operator action is modeled in the PRA as OPRHC1 (for small LOCAs) and OPRHM1 (for medium LOCAs); the only difference between these is in the required timing of the action based on the bounding leak rate for the LOCA category, and in the resulting HEPs. Large LOCAs in the PRA model do not credit HHSI flow to prevent or mitigate core damage.

For a large LOCA, this path is modeled only to determine the extent of the RWST inventory available in containment after core damage occurs, for use in the Level 2 model. The nominal HEPs for these operator actions (OPRHC1 and OPRHM1, and all hazard-specific versions) are used in this assessment, despite the fact that dedicated operators will be stationed both at the benchboard control for MOV-1SI-836 and at the physical location of the valve itself in order to ensure the valve will be quickly opened if a Safety Injection is required. While the presence of the dedicated operators will greatly improve the likelihood of successfully opening the valve, the baseline risk assessment conservatively does not take explicit credit for this compensatory measure. Improved HEPs are credited in sensitivity cases only. Also of note is that these operator actions only credit operation of the valve from the benchboard control; the PRA does not model a local action to open MOV-1SI-836.

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 6 of 24 Once these basic event values were set to 1.0, all top events in which they are used were re-quantified using the BDD split fraction quantification:

HH, HC, LO, LC, HM, LQ, LM, HL, XL, LL, OA

A new Master Frequency File (MFF) was created, named REV8MFFS, and the event trees were re-quantified for both CDF and LERF. A sequence truncation limit of 1E-14 was used in the event tree quantification.

With the normal Safety Injection flowpath isolated, the configuration CDF is determined to be 1.85E-04 /yr. The BV1REV8 CDF is 6.31E-05 /yr; therefore the delta CDF of the proposed configuration to support this one-time TS CT change is (1.85E-04 - 6.31E-05) = 1.22E-04 /yr. Dividing this delta CDF by 8,760 hrs/yr to obtain the conditional CDF yields (1.22E-04 / 8760) = 1.39E-08 /hr. Co mparing this to the ICCDP acceptance criteria of 1.0E-06 (1.0E-06 / 1.39E-08 /hr), it is determined that the proposed configuration may remain in effect for up to 72.01 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months before reaching the ICCDP acceptance criteria of Regulatory Guide 1.177. For the requested CT of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , the ICCDP is 5.00E-07.

Similarly, the configuration LERF is determined to be 2.05E-06 /yr. The BV1REV8 LERF is 1.12E-06 /yr; therefore the delta LERF of the proposed configuration to support this one-time TS CT change is (2.05E 1.12E-06) = 9.36E-07 /yr. Dividing this delta LERF by 8,760 hrs/yr to obtain the conditional LERF yields (9.36E-07 / 8760)

= 1.07E-10 /hr. Comparing this to the ICLERP acceptance criteria of 1.0E-07 (1.0E-07 / 1.07E-10 /hr), it is determined that the proposed configuration may remain in effect for up to 936.11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months before reaching the ICLERP acceptance criteria of Regulatory Guide 1.177. For the requested CT of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , the ICLERP is 3.85E-09.

These results can be seen in Attachment 1.

Top Risk Contributors (RG 1.177 Tier 2)

For the proposed configuration, 302 initiating events (approximately 39% of the initiating events in the model) show an increase of greater than 10% from their nominal CDF. As expected, these are initiating events in which a LOCA occurs, whether as the initiating event itself, or as a consequential failure occurring following the initiating event. A listing of these initiating events, including their total CDF and percent increases from their nominal CDF, is provided in Attachment 2. This evaluation of the top risk contributors focuses on CDF and not LERF since the ICCDP is the limiting metric for this assessment, compared to ICLERP, and because a review of LERF results showed no outliers with respect to the results for CDF.

Overall the risk increase associated with this proposed plant configuration is dominated by fire scenarios in which a LOCA may result, whether by spurious opening of a PORV, 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 re-close, failure of RCP seal injection and shutdown seals (SDS) with a resultant RCP seal LOCA, etc.

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 7 of 24 The most significant CDF increases, when considering both relative change and absolute resulting CDF, are determined to be from numerous fire scenarios in fire compartments 1-CR-4 and 1-ES-1.

1-CR-4 is the Process Instrumentation room, containing both safety trains of SSPS and Reactor Protection cabinets, and is the single most risk significant fire compartment in the Fire PRA. Most circuits associated with control room instruments and controls have cable routing through 1-CR-4. The fire scenario in 1-CR-4 showing the highest CDF for this configuration is FCR4F2, a fire occurring in sections 5, 8-13, 19, 21-22, 24-29 of the Primary Process Racks. Pressurizer pressure transmitters which provide automatic control signals for the PORVs are affected by the fire, which also causes a spurious Safety Injection signal resulting in a challenge to the primary safety valves which then fail to re-close, requiring Safety Injection flow to mitigate the loss of inventory. The modeling in this case is conservative since the spurious SI signal will not automatically open MOV-1SI-836 and the normal SI flowpath will be isolated; however it is possible that the initial SI signal may not be recognized as spurious and the dedicated operator would open MOV-1SI-836, so the modeling was not altered. The most likely core damage progression from this point in the scenario is a failure of the operators to initiate cooldown and depressurization, which prevents use of the SI accumulators and LHSI to mitigate the LOCA.

1-ES-1 is the Train A Emergency Switchgear room. All fires in this room are assumed to result in failure of the Train A Emergency Bus. Power to MOV-1SI-836 is provided from the Train A Emergency bus via MCC-1-E5, so HHSI flow is modeled as being unavailable for any fire in 1-ES-1 which may occur while the plant remains in the proposed configuration. This modeling is conservative because not all fires will realistically fail the Train A Emergency bus, and because no credit is taken for the dedicated operator stationed locally at MOV-1SI-836 to open the valve if power is not available. The fire scenario in 1-ES-1 showing the highest CDF for this configuration is FES1B7, a fire occurring in 480V transformer TRANS-1-8N. The fire causes failure of Vital Bus Red (1) and Vital Bus Blue (3), which creates a Safety Injection signal due to loss of power to Train A SSPS. The HHSI flow is assumed to demand primary pressure relief from the PORVs, but the PORVs fail to open due to loss of Vital Bus power and other fire impacts, challenging the primary safety valves which subsequently fail to re-close and require HHSI flow to make up the inventory loss.

Similar to the scenario discussed for 1-CR-4, this modeling is conservative since the spurious SI signal will not automatically open MOV-1SI-836 and the normal SI flowpath will be isolated; however it is possible that the initial SI signal may not be recognized as spurious and the dedicated operator would open MOV-1SI-836, so the modeling was not altered. The most likely core damage progression from this point in the scenario again involves failure of the operators to initiate cooldown and depressurization, which prevents use of the SI accumulators and LHSI to mitigate the LOCA.

The small LOCA initiating event SLOCN also exhibited a significant risk increase and is among the top risk contributors in the proposed configuration. Failure of the operator action to open MOV-1SI-836 is the most prevalent factor driving the core damage PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 8 of 24 sequences for this initiating event. As described above, however, this assessment takes no credit for the compensatory measure of stationing dedicated operators to ensure this valve will open. With a dedicated operator stationed both at the benchboard and at the valve itself it can be assumed that this action would be more successful than its nominal value, and later sensitivity cases prove this to be the case by showing the initiating event contribution of SLOCN to be substantially lowered.

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). Seismic and flooding scenarios are not significantly affected by the proposed plant configuration and do not contribute substantially to the increased risk.

Configuration Risk Considerations (RG 1.177 Tier 3)

Configuration risk is managed in accordance with 10CFR50.65(a)(4), using Phoenix Risk Monitor to track planned and emergent work to determine impact to configuration risk. This will ensure that potentially lower probability, but nonetheless risk-significant, configurations resulting from maintenance and other operational activities will be identified and compensated for while performing this repair.

Furthermore, while the plant remains in this proposed configuration only one additional planned maintenance activity will be performed on PRA components, in order to minimize the risk impact of this repair evolution.

The additional planned maintenance activity intended to be performed during this proposed repair configuration is cleaning of Intake Bay C, which renders River Water pump 1C (1WR-P-1C) unavailable. Failure / unavailability of 1WR-P-1C was not identified as a significant risk contributor or a high risk configuration in Tiers 1 and 2, so it falls into Tier 3 for assessment. A configuration-specific case was run in the Riskman Mini-Monitor confirming that this activity does not represent a significant increase in configuration risk; therefore this activity will be managed by the normal station 10CFR50.65(a)(4) program and work control process.

Sensitivity Cases

Additional sensitivity cases were considered to address the impacts of potential conservatisms in the analysis, and to determine a bounding worst case for the proposed configuration.

o Sensitivity Case 1

This sensitivity case takes credit for the compensatory measure of stationing dedicated operators both at the benchboard control for MOV-1SI-836 and at the valve itself to ensure it will open if a Safety Injection is required while the normal SI flowpath is isolated, by guaranteeing success of the action to open the valve. This is not considered a realistic case; the realistic case would be PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 9 of 24 somewhere between this sensitivity and the baseline case using the nominal HEP values for the operator action. This represents a best-case scenario for the HEP. The application model B1R8HCL1 created to evaluate the proposed repair configuration as described above is used as the starting point for this sensitivity. B1R8HCL1 is cloned and renamed as B1R8HCS1, and the following additional basic event changes are made to guarantee success of the operator action (and the hazard-specific versions):

OPRHC1=0.0 OPRHC1F1=0.0 OPRHC1F2=0.0 OPRHC1F3=0.0 OPRHC1S1=0.0 OPRHC1S2=0.0 OPRHC1S3=0.0 OPRHM1=0.0

The delta CDF for this sensitivity case is determined to be 7.89E-05, which would permit this configuration to remain in effect for up to 111.05 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months before reaching the ICCDP acceptance criteria of 1.0E-06. The proposed CT time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months results in an ICCDP of 3.24E-07.

The delta LERF for this sensitivity case is determined to be 5.50E-07, which would permit this configuration to remain in effect for up to 1592.43 hours4.976852e-4 days 0.0119 hours 7.109788e-5 weeks 1.63615e-5 months before reaching the ICLERP acceptance criteria of 1.0E-07. The proposed CT time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months results in an ICLERP of 2.26E-09.

o Sensitivity Case 2

This sensitivity case takes more realistic credit for the compensatory measure of dedicating operators both at the benchboard control for MOV-1SI-836 and at the valve itself to ensure it will open if a Safety Injection is required while the normal SI flowpath is isolated, by lowering the failure probability of the action to open the valve. This is considered a more realistic representation than Sensitivity Case 1; this case changes the HEP of these actions to a value of 1.0E-04. This represents a reduction from the nominal HEPs used in the base case (nominal OPRHC1 = 2.10E-03; OPRHM1 = 1.20E-02), to allow some additional credit for the dedicated operators. The application model B1R8HCL1 created to evaluate the proposed repair configuration as described above is used as the starting point for this sensitivity. B1R8HCL1 is cloned and renamed as B1R8HC4, and the following additional basic event changes are made to reduce the credited failure probability of the operator action to open MOV-1SI-836 (and the hazard-specific versions):

OPRHC1=1.0E-04 OPRHC1F1=1.0E-04 OPRHC1F2=1.0E-04 PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 10 of 24 OPRHC1F3=1.0E-04 OPRHC1S1=1.0E-04 OPRHC1S2=1.0E-04 OPRHC1S3=1.0E-04 OPRHM1=1.0E-04

The delta CDF for this sensitivity case is determined to be 7.90E-05, which would permit this configuration to remain in effect for up to 110.94 hours0.00109 days 0.0261 hours 1.554233e-4 weeks 3.5767e-5 months before reaching the ICCDP acceptance criteria of 1.0E-06. The proposed CT time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months results in an ICCDP of 3.25E-07.

The delta LERF for this sensitivity case is determined to be 5.53E-07, which would permit this configuration to remain in effect for up to 1583.67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months before reaching the ICLERP acceptance criteria of 1.0E-07. The proposed CT time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months results in an ICLERP of 2.27E-09.

o Sensitivity Case 3

This sensitivity case investigates the result of assuming a guaranteed failure of MOV-1SI-836 to open while in the proposed repair configuration. This is considered to represent an extremely conservative and unrealistic worst case scenario. This sensitivity is performed to determine bounding risk values for the proposed configuration. The application model B1R8HCL1 created to evaluate the proposed repair configuration as described above is used as the starting point for this sensitivity. B1R8HCL1 is cloned and renamed as B1R8HCS5, and the following additional model changes are made to guarantee failure of MOV-1SI-836:

MVFOMOVSI836=1.0 and Remove the appropriate common cause group including MOV-1SI-867C and MOV-1SI-867D from top events HH (FMOVFO), HC (EMOVO), LO (CMOVO), HL (EMOVO), HM (EMOVO), LQ (CMOVO), XL (CMOVO), and OA (FMOVO). (Note that quantification also requires top events LC, LM, and LL in order to account for conditional split fractions.) This prevents the CCF value less than 1.0 for MOV-1SI-836 from being used instead of the total failure rate set by the normal basic event for the valve.

The delta CDF for this bounding sensitivity case is determined to be 9.11E-04, which would only permit this configuration to remain in effect for up to 9.61 hours7.060185e-4 days 0.0169 hours 1.008598e-4 weeks 2.32105e-5 months before reaching the ICCDP acceptance criteria of 1.0E-06. The proposed CT time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months results in an ICCDP of 3.74E-06.

The delta LERF for this bounding sensitivity case is determined to be 3.12E-05, which would only permit this configuration to remain in effect for up to 28.11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months before reaching the ICLERP acceptance criteria of 1.0E-07. The proposed CT time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months results in an ICLERP of 1.28E-07.

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 11 of 24

The results of all sensitivity cases are shown in Attachment 3.

7.0 ASME PRA Standard Evaluation

The BVPS-1 PRA (PRA-BV1-AL-R08) is an at-power, integrated Level 2 PRA that includes internal (including internal floods) and external events (fires and seismic). Other external hazards were screened from the PRA model, and due to the nature of these external hazards not directly causing a LOCA there is nothing about the proposed configuration which would significantly increase the risk of other external hazards to the point of becoming important to this assessment. The model does not include shutdown events.

The BVPS-1 PRA internal events, fire, and seismic external events models satisfy the guidance provided in Regulatory Guide 1.200, Revision 2, and meets Capability Category II with all Facts and Observations (F&Os) addressed.

The BVPS-1 seismic PRA model (SPRA) has been peer reviewed against the requirements of the ASME/ANS PRA Standard (RA-Sb-2013) and RG 1.200, Revision 2 through the gap assessment between Addendum A and B of the ASME/ANS PRA Standard.

The Fire PRA model was upgraded to the state-of-the art in order to support the NFPA 805 fire protection licensing basis and is in compliance with approved NRC guidance issued in support of NFPA 805 and the ASME PRA Standard RA-Sa-2009, as reflected in the BVPS NFPA 805 Safety Evaluation Report (SER) issued by the NRC.

8.0 Results

The PRA quantitative and qualitative analyses support this one-time CT change to Technical Specification 3.5.2. The ICCDP for the requested 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months CT is 5.00E-07, which is significantly less than the ICCDP acceptance criterion of 1.0E-06 as described in Regulatory Guide 1.177. Likewise the ICLERP for the requested 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months CT is 3.85E-09, which is also significantly less than the ICLERP acceptance criterion of 1.0E-07 as described in Regulatory Guide 1.177.

In consideration of defense in depth, due to the significant contribution to the risk increase of certain fire scenarios as described above in Section 6.0 of this assessment, it is recommended that a continuous fire watch be established and no hot work will be permitted in fire compartments 1-CR-4 and 1-ES-1 for the duration of this repair activity, as an additional compensatory measure. This will help to reduce the actual risk of this configuration by allowing for personnel to promptly extinguish any fire that may ignite before it can cause significant damage.

It is also recommended that control room operators will be briefed on the importance of beginning to cooldown and depressurize in accordance with the EOP Network if MOV-1SI-836 should fail to open, in order to allow injection from the SI Accumulators and ultimately LHSI to mitigate the loss of inventory.

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 12 of 24 Attachment 1

Results

Attachment1AICCDP BV1REV8 B1R8HCL1 Proposed Group EffectiveModel Configuration FireCDF: 4.84E05 1.67E04 FloodCDF: 5.72E07 5.84E07 InternalCDF(includesFlood): 2.76E06 6.03E06 SeisCDF: 1.19E05 1.21E05 Total CDF 6.31E05 1.85E04 DeltaCDF= (1.85E04) (6.31E 05) = 1.22E04 /yr ConditionalCDF= (1.22E04) /8760= 1.39E08 /hr TimetoreachICCDP Limit(1.0E06) = (1.0E06) /(1.39E08) = 72.01 hrs ICCDPforrequested 36hours= (1.39E08) *36 = 5.00E07

Attachment1B ICLERP BV1REV8 B1R8HCL1 Proposed Group EffectiveModel Configuration InternalLERF(includesFlood): 1.05E07 2.40E07 FireLERF: 4.22E07 1.22E06 SeisLERF: 5.90E07 5.90E07 FloodLERF: 2.38E09 2.38E09 Total LERF 1.12E06 2.05E06 DeltaLERF= (2.05E06) (1.12E 06) = 9.36E07 /yr ConditionalLERF = (9.36E07) /8760= 1.07E10 /hr TimetoreachICLERPLimit (1.0E07) = (1.0E07) /(1.07E10) = 936.11hrs ICLERPforrequested36 hours = (1.07E10) *36 = 3.85E09

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 13 of 24 Attachment 2

Risk Contributors

B1R8HCL1 BV1REV8 Results Results Comparison IE %Changein Initiator Type DescriptionCDF CDFCDF FES1B7 FIRE TRANS1 8N FDS1 9.47E06 3.01E07 3041.89%

FCR4F2 FIRE RKPRI PROC (5, 813, 19, 2122, 2429) FDS0 8.44E06 6.91E07 1121.59%

FMCR77 FIRE DC3JB501 4.33E06 2.98E07 1351.88%

FES136 FIRE MCC1 E9 FDS0 3.66E06 1.12E07 3168.58%

FCR4F4 FIRE RKPRI PROC (5, 813, 19, 2122, 2429) FDS2 3.40E06 2.17E07 1462.63%

ComputerCabinetsRKCMPDIN 4, TERM2, FCR472 FIREIPCCAB05, RKCMP TERM1 FDS1 3.15E06 1.88E07 1574.15%

FCS1T1 FIRE 1CS 1BIN18 TS#221 2.82E06 8.34E08 3281.18%

FES164 FIRE 480VUS1 8 N(Sections 111)FDS2 2.80E06 2.36E07 1087.84%

FCR4V2 FIRE RKREACPROT A FDS1A 2.77E06 1.44E07 1829.35%

FNS1F4 FIRE 4KVS1D 1, 2, 3, 4, 4A(HEAF)FDS2 2.63E06 1.89E06 39.02%

IE SLOCN LOCA SMALL LOCA,NONISOLABLE 2.33E06 4.57E07 410.24%

FES168 FIRE 480VUS1 8 N(Sections1222)FDS2 2.24E06 1.89E07 1087.90%

FCV160 FIRE TS#7FDS1/2/5 2.14E06 4.74E07 350.92%

FES182 FIRE 4160VBUS4KVS1AE(Sec1,2,3,4,5) FDS2 2.04E06 1.62E07 1162.45%

FCR468 FIRE RKSEC PROC B,D,H,J,K,L,M FDS2 1.96E06 2.76E07 609.77%

FCR4I9 FIRE RKVV RELA FDS1 1.75E06 7.19E08 2339.17%

ComputerCabinetsRKCMPDIN 4, TERM2, FCR473 FIREIPCCAB05, RKCMP TERM1 FDS2 1.73E06 1.78E07 869.10%

FES107 FIRE BATCHG1 1 FDS1 1.69E06 1.42E07 1087.98%

FES194 FIRE 4160VBUS4KVS1AE(S14,15,16,17) FDS2 1.66E06 1.39E07 1088.02%

FES190 FIRE 4160VBUS4KVS1AE(S10,11,12,13) FDS2 1.65E06 1.39E07 1088.20%

FCR493 FIRE PNLREL 43 FDS2 1.65E06 2.32E07 609.81%

FNS1E2 FIRE 4KVS1D 1, 2, 3, 4, 4AFDS2 1.65E06 1.18E06 39.02%

FES186 FIRE 4160VBUS4KVS1AE(Sec7,8,8A,9) FDS2 1.63E06 1.29E07 1162.54%

FCR4J6 FIRE RKREACPROT B FDS2 1.54E06 1.75E07 780.12%

FCV117 FIRE MCC1E5 FDS1 1.51E06 2.86E07 429.35%

FES198 FIRE 4160VBUS4KVS1AE(1,2,3,4,5)HEAF FDS2/5 1.48E06 1.17E07 1162.59%

FCR4V0 FIRE RKPRI PROC (5, 813, 19, 2122, 2429) FDS1A 1.39E06 1.85E07 652.42%

FCR4N1 FIRE RKANN DMX FDS2 1.38E06 3.74E08 3590.96%

FCR4J5 FIRE RKREACPROT B FDS1 1.38E06 1.67E07 723.78%

RKPRI PROC 1, 2, 3, 6, 7, 4, 20, 23, 14, 15, 16, FCR4F8 FIRE 17, 18 120VACPrimaryFDS0 1.27E06 8.19E07 55.12%

FCR4F3 FIRE RKPRI PROC (5, 813, 19, 2122, 2429) FDS1 1.25E06 1.05E07 1096.86%

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 14 of 24

B1R8HCL1 BV1REV8 Results Results Comparison IE %Changein Initiator Type DescriptionCDF CDFCDF FCR423 FIRE UPSCMP1 FDS1 1.24E06 1.28E08 9565.20%

FES1B3 FIRE 4160VBS4KV1AE(14,15,16,17)HEAF FDS2/5 1.20E06 1.01E07 1088.14%

4160VBUS4KVS1AE(10,11,12,13) HEAF FES1A8 FIRE FDS2/5 1.20E06 1.01E07 1088.14%

FES1A3 FIRE 4160VBUS4KVS1AE(7,8,8A,9)HEAF FDS2/5 1.18E06 9.37E08 1162.70%

FES1C1 FIRE Bin18 1.16E06 2.37E08 4791.27%

FES111 FIRE BATCHG1 3 FDS2 1.12E06 8.83E08 1162.73%

FCR4P8 FIRE RK3OVAPROC 20 FDS2 1.11E06 4.80E07 131.86%

FCR4P3 FIRE RK1OVAPROC 10 FDS2 1.11E06 4.80E07 131.86%

FES132 FIRE INVVITBUS1 3FDS0 1.10E06 2.25E08 4792.32%

FCR492 FIRE PNLREL 43 FDS1 1.08E06 1.52E07 610.47%

FCR463 FIRE RKSEC PROC A,C,E,F,G,P FDS3 1.06E06 1.50E07 609.92%

FCR4A7 FIRE RKCMPDIN 5FDS1 1.02E06 1.27E07 704.18%

FCR4A2 FIRE RKCMPTERM3,4 FDS1 1.02E06 2.85E07 257.33%

FCR4G6 FIRE RKREACPROT A FDS2 1.01E06 6.80E08 1383.57%

FNS1J0 FIRE TRANS1 4G FDS2 9.56E07 6.75E07 41.72%

FCV118 FIRE MCC1E5 FDS2 9.51E07 1.78E07 433.19%

FES199 FIRE 4160VBUS4KVS1AE(1,2,3,4,5)HEAF FDS3/6 9.41E07 7.46E08 1162.83%

FCR4K6 FIRE RKREACPROT B FDS1A 9.26E07 1.13E07 721.62%

FCR462 FIRE RKSEC PROC A,C,E,F,G,P FDS2 9.01E07 4.56E08 1878.63%

FCR4N0 FIRE RKANN DMX FDS1 8.99E07 1.69E08 5227.83%

FCR4V3 FIRE RKREC P TST A FDS1A 8.64E07 4.47E08 1832.67%

FCR4H1 FIRE RKREC P TST A FDS1 8.47E07 4.10E08 1965.62%

FCR4C3 FIRE DL1RC 100A ICCMonitorTrainA &BFDS2 8.45E07 1.14E07 641.07%

FCV113 FIRE MCC1E11 FDS2 8.37E07 1.57E07 433.27%

FES177 FIRE 480VUS1 8 N(Sections1222)HEAF FDS2/5 8.19E07 6.89E08 1088.28%

FES129 FIRE INVVITBUS1 1FDS1 8.00E07 1.79E08 4375.35%

FES1B4 FIRE 4160VBS4KV1AE(14,15,16,17)HEAF FDS3/6 7.63E07 6.42E08 1088.36%

4160VBUS4KVS1AE(10,11,12,13) HEAF FES1A9 FIRE FDS3/6 7.63E07 6.42E08 1088.36%

FES118 FIRE BATBKR1 1 FDS1 7.57E07 6.37E08 1088.50%

FCS1O8 FIRE BIN18TS#1611TC589O 7.56E07 6.13E08 1132.52%

FES1A4 FIRE 4160VBUS4KVS1AE(7,8,8A,9)HEAF FDS3/6 7.53E07 5.96E08 1162.98%

FCS1I0 FIRE TS#27BBIN6/7 FDS1 7.44E07 1.80E07 312.28%

FCV119 FIRE MCC1E5 FDS3 7.02E07 1.32E07 433.38%

FCR424 FIRE UPSCMP1 FDS2 6.99E07 2.67E07 162.35%

FCR4X6 FIRE 1CR 4RK VV RELBFDS1A 6.75E07 7.94E08 749.61%

FCR4V4 FIRE RKREACPROT B FDS1B 6.69E07 8.32E08 704.00%

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 15 of 24

B1R8HCL1 BV1REV8 Results Results Comparison IE %Changein Initiator Type DescriptionCDF CDFCDF FES126 FIRE BATBKR1 3 FDS2 6.65E07 5.27E08 1163.06%

FES122 FIRE DCSWBD13 FDS2 6.65E07 5.27E08 1163.06%

FCR4E4 FIRE RKANN FDS3 6.60E07 2.84E07 131.87%

FCR4F5 FIRE RKPRI PROC (5, 813, 19, 2122, 2429) FDS3 6.57E07 5.54E08 1086.29%

FNS1N0 FIRE 1NS14KVS 1D1 2344AFDS2b 6.54E07 4.69E07 39.34%

FCS1O9 FIRE BIN18TS#1951TC413O 6.38E07 5.15E08 1138.64%

FCR4N6 FIRE TurbineControlCabinet(RKEHC) FDS2 6.06E07 8.39E08 622.85%

FES178 FIRE 480VUS1 8 N(Sections1222)HEAF FDS3/6 5.98E07 5.03E08 1088.46%

FCR4G7 FIRE RKREACPROT A FDS3 5.97E07 6.60E08 805.43%

FNS1I9 FIRE TRANS1 4G FDS1 5.96E07 4.24E07 40.56%

FCR4O8 FIRE IPCRK SDP FDS3 5.77E07 2.50E07 130.50%

FCR4B8 FIRE SER(RKSEQ) FDS2 5.63E07 2.32E07 142.58%

FCR457 FIRE CommunicationsBatteryCharger48BFDS2 5.59E07 9.16E08 510.51%

FCS1K8 FIRE TS#31BIN6/7 FDS1 5.49E07 4.41E08 1144.23%

FCV116 FIRE MCC1E5 FDS0 5.44E07 6.96E08 681.13%

FES128 FIRE INVVITBUS1 1FDS0 5.32E07 1.09E08 4795.36%

FCR4A3 FIRE RKCMPTERM3,4 FDS2 5.22E07 7.04E08 641.29%

FCR4F6 FIRE RKPRI PROC (5, 813, 19, 2122, 2429) FDS4 5.21E07 5.12E08 917.71%

FCV157 FIRE TS#6FDS1/2/5 5.03E07 9.42E08 433.88%

FCS1O7 FIRE BIN18TS#1091TC412O 4.91E07 3.96E08 1141.08%

FCR4D8 FIRE PNLREL 42 FDS2 4.81E07 6.48E08 641.33%

FCR4A8 FIRE RKCMPDIN 5FDS2 4.51E07 6.08E08 641.36%

FES152 FIRE Relay Panel35FFDS2 4.20E07 1.33E08 3046.88%

FDG121 FIRE 1EE EG1 FDS0 4.05E07 1.66E07 144.36%

RKPRI PROC 1, 2, 3, 6, 7, 4, 20, 23, 14, 15, 16, FCR4V1 FIRE 17, 18 120VACPrimaryFDS1A 3.96E07 1.11E07 257.13%

MLOCA IE MEDIUMLOSSOFCOOLANTACCIDENTIN C LOCA LOOPC 3.93E07 1.04E07 278.24%

MLOCA IE MEDIUMLOSSOFCOOLANTACCIDENTIN B LOCA LOOPB 3.93E07 1.04E07 278.24%

MLOCA IE MEDIUMLOSSOFCOOLANTACCIDENTIN A LOCA LOOPA 3.93E07 1.04E07 278.24%

FES163 FIRE 480VUS1 8 N(Sections 111)FDS1 3.89E07 1.59E08 2347.53%

FES140 FIRE PNLDGSEQ 1 FDS2 3.82E07 1.21E08 3047.22%

FCR4M7 FIRE PNLV/LP MON FDS3 3.76E07 1.43E07 162.39%

FES134 FIRE INVVITBUS1 3FDS2 3.76E07 2.97E08 1163.58%

FCR4E9 FIRE RKCMPDIN 6FDS2 3.75E07 1.62E07 131.89%

RKPRI PROC 1, 2, 3, 6, 7, 4, 20, 23, 14, 15, 16, FCR4G2 FIRE 17, 18 120VACPrimaryFDS4 3.73E07 2.48E08 1403.17%

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 16 of 24

B1R8HCL1 BV1REV8 Results Results Comparison IE %Changein Initiator Type Description CDF CDF CDF FCR4D3 FIRE PNLAMSAC FDS2 3.69E07 5.20E08 610.36%

FCR4D2 FIRE PNLAMSAC FDS1 3.45E07 4.86E08 610.95%

FCR4O1 FIRE PNLSHUTDOWN FDS2 3.33E07 1.31E07 155.06%

FES167 FIRE 480VUS1 8 N(Sections1222)FDS1 3.11E07 1.27E08 2347.95%

FCR4O2 FIRE PNLSHUTDOWN FDS3 3.10E07 1.61E07 93.13%

FCR4J0 FIRE RKVV RELA FDS2 2.89E07 3.23E08 794.67%

FCR4O7 FIRE IPCRK SDP FDS2 2.67E07 1.05E07 155.07%

FMCR68 FIRE DC3 2.55E07 2.10E08 1117.50%

FES156 FIRE Relay Panel37FFDS2 2.54E07 8.07E09 3048.93%

FES148 FIRE Relay Panel33FFDS2 2.54E07 8.07E09 3048.93%

FES144 FIRE Relay Panel31FFDS2 2.54E07 8.07E09 3048.93%

FCR4H2 FIRE RKREC P TST A FDS2 2.51E07 2.57E08 875.94%

FCR4L3 FIRE RKAUX RPTSTB FDS1 2.47E07 2.90E08 750.58%

FCR4K7 FIRE RKAUX REL BFDS1 2.47E07 2.09E08 1079.52%

FCR4K1 FIRE RKREC P TST B FDS1 2.46E07 2.90E08 747.60%

FCR4D7 FIRE PNLREL 42 FDS1 2.44E07 3.02E08 705.31%

FCR4I4 FIRE RKAUX RPTSTA FDS2 2.33E07 2.61E08 794.92%

FCR4H8 FIRE RKAUX REL A FDS2 2.33E07 2.61E08 793.91%

FES1A0 FIRE 4160VBUS4KVS1AE(1,2,3,4,5)HEAF FDS4/7 2.27E07 1.90E08 1089.35%

RKPRI PROC 1, 2, 3, 6, 7, 4, 20, 23, 14, 15, 16, FCR4G0 FIRE 17, 18 120VACPrimaryFDS2 2.24E07 7.85E08 184.62%

FDG132 FIRE 1DG 1TS#8 2.22E07 9.07E08 144.38%

FCR4E3 FIRE RKANN FDS2 2.21E07 1.69E07 30.41%

FES115 FIRE DCSWBD11 FDS2 2.16E07 1.82E08 1089.31%

IE INTERFACINGSYSTEMSLOCA(VSEQUENCE VSX LOCA BDDMC) 2.10E07 8.00E08 162.23%

FCR4E2 FIRE RKANN FDS1 2.01E07 1.22E07 64.02%

FES160 FIRE SW1 8N1 FDS2 1.99E07 6.32E09 3050.04%

FCR4C2 FIRE DL1RC 100A ICCMonitorTrainA &BFDS1 1.97E07 2.44E08 705.45%

FCR4P7 FIRE RK3OVAPROC 20 FDS1 1.88E07 5.83E08 221.49%

FCR4P2 FIRE RK1OVAPROC 10 FDS1 1.88E07 5.83E08 221.49%

FES1B5 FIRE 4160VBS4KV1AE(14,15,16,17)HEAF FDS4/7 1.81E07 1.52E08 1089.63%

4160VBUS4KVS1AE(10,11,12,13) HEAF FES1B0 FIRE FDS4/7 1.81E07 1.52E08 1089.63%

FES1A5 FIRE 4160VBUS4KVS1AE(7,8,8A,9)HEAF FDS4/7 1.81E07 1.52E08 1089.63%

FES1D8 FIRE TS#17 1.79E07 5.69E09 3050.24%

FRC125 FIRE 1RCP 1B FDS0 1.75E07 1.32E07 32.96%

FRC122 FIRE 1RCP 1A FDS0 1.75E07 1.32E07 32.96%

FRC128 FIRE 1RCP 1C FDS0 1.75E07 1.32E07 32.96%

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 17 of 24

B1R8HCL1 BV1REV8 Results Results Comparison IE %Changein Initiator Type Description CDF CDF CDF FCR4N7 FIRE TurbineControlCabinet(RKEHC) FDS3 1.68E07 6.41E08 162.48%

FCR4U0 FIRE 1CR 4TS#20 Bins6&7 1.56E07 1.36E08 1048.90%

FCR4G4 FIRE RKREACPROT A FDS0 1.50E07 1.22E07 22.71%

FES130 FIRE INVVITBUS1 1FDS2 1.48E07 1.24E08 1089.78%

FCR4G5 FIRE RKREACPROT A FDS1 1.48E07 1.26E07 17.26%

FES179 FIRE 480VUS1 8 N(Sections1222)HEAF FDS4/7 1.31E07 1.10E08 1090.06%

FDG108 FIRE EEC 2A FDS0 1.31E07 5.34E08 144.40%

FCS1H7 FIRE TS#27ABIN6/7FDS1 1.25E07 2.91E08 329.13%

FCV120 FIRE MCC1E5 FDS4 1.23E07 2.28E08 439.08%

FES172 FIRE 480VUS1 8 N(Sections111)HEAFFDS2/5 1.14E07 9.55E09 1090.15%

FDG111 FIRE EEC 1A FDS0 1.09E07 4.45E08 144.41%

IE POW AOX ER LOSSOFEMERGENCY4160VACORANGE 1.07E07 3.20E08 233.16%

FRC120 FIRE 1RSP 1B FDS1 1.04E07 7.83E08 33.03%

FRC118 FIRE 1RSP 1A FDS1 1.04E07 7.83E08 33.03%

FRC112 FIRE 1DAP 4B FDS1 1.04E07 7.83E08 33.03%

FRC110 FIRE 1DAP 4A FDS1 1.04E07 7.83E08 33.03%

FES137 FIRE MCC1 E9 FDS1 1.03E07 8.63E09 1090.44%

FCR464 FIRE RKSEC PROC A,C,E,F,G,P FDS4 9.51E08 1.34E08 611.45%

FCR469 FIRE RKSEC PROC B,D,H,J,K,L,M FDS3 9.39E08 1.32E08 611.46%

FCR445 FIRE BAT1 5 FDS2 9.17E08 5.36E08 71.15%

FCS124 FIRE 1CS 1BIN5 TS#221 8.88E08 7.48E09 1086.45%

FDG122 FIRE 1EE EG1 FDS1 8.85E08 4.27E08 107.31%

FCR425 FIRE UPSCMP1 FDS3 8.71E08 3.32E08 162.58%

FDG129 FIRE TS#5 8.58E08 3.51E08 144.42%

FCV112 FIRE MCC1E11 FDS1 8.45E08 6.70E08 26.10%

FES173 FIRE 480VUS1 8 N(Sections111)HEAFFDS3/6 8.30E08 6.97E09 1090.91%

RKPRI PROC 1, 2, 3, 6, 7, 4, 20, 23, 14, 15, 16, FCR4F9 FIRE 17, 18 120VACPrimaryFDS1 8.06E08 5.21E08 54.85%

RKPRI PROC 1, 2, 3, 6, 7, 4, 20, 23, 14, 15, 16, FCR4G1 FIRE 17, 18 120VACPrimaryFDS3 7.70E08 3.46E09 2122.61%

FRC130 FIRE Bin18 7.58E08 6.04E08 25.46%

FRC116 FIRE 1RHP 1B FDS1 7.57E08 5.69E08 33.03%

FRC114 FIRE 1RHP 1A FDS1 7.57E08 5.69E08 33.03%

FRC105 FIRE 1VSF1C FDS0 7.43E08 5.59E08 32.96%

FRC108 FIRE 1VSF2C FDS0 7.43E08 5.59E08 32.96%

FRC106 FIRE 1VSF2A FDS0 7.43E08 5.59E08 32.96%

FRC104 FIRE 1VSF1B FDS0 7.43E08 5.59E08 32.96%

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 18 of 24

B1R8HCL1 BV1REV8 Results Results Comparison IE %Changein Initiator Type Description CDF CDF CDF FRC103 FIRE 1VSF1A FDS0 7.43E08 5.59E08 32.96%

FRC107 FIRE 1VSF2B FDS0 7.43E08 5.59E08 32.96%

FCR4K0 FIRE RKANN FDS1A 7.39E08 1.30E08 467.58%

FCS123 FIRE 1CS 1BIN5 TS#161 7.33E08 6.19E09 1084.00%

FCR461 FIRE RKSEC PROC A,C,E,F,G,P FDS1 7.23E08 2.91E08 148.13%

IE POW LOSPG ER LOSSOFOFFSITEPOWERGRIDCENTERED 6.56E08 1.49E08 339.03%

FCR4V5 FIRE RKREC P TST B FDS0A 6.45E08 3.95E08 63.30%

ComputerCabinetsRKCMPDIN 4, TERM2, FCR474 FIREIPCCAB05, RKCMPTERM1 FDS3 6.42E08 7.38E09 770.11%

FCR4E5 FIRE RKANN FDS4 6.40E08 2.76E08 132.03%

FCS1S8 FIRE 1CS 1BIN5 TS#195 6.19E08 5.21E09 1088.61%

FES205 FIRE TRF1P15 FDS0 6.09E08 3.98E08 52.93%

FES280 FIRE 480VUS1 9 FDS3 5.89E08 4.97E08 18.62%

FCR439 FIRE RKPRI PROC (3037)FDS3 5.63E08 4.96E09 1033.60%

IE TRAN TTRIP S TURBINETRIP 5.50E08 1.19E08 362.68%

FCS1J5 FIRE 1CS 1TS#29A BIN6/7FDS1 5.41E08 5.26E09 928.50%

FDG126 FIRE TS#2 4.75E08 1.95E08 144.28%

FCR494 FIRE PNLREL 43 FDS3 4.72E08 6.62E09 612.40%

FCR4O9 FIRE 1CR 4IPC RK SDPFDS4 4.67E08 2.03E08 130.67%

IE SLOCI LOCA SMALL LOCA,ISOLABLE 4.61E08 8.56E10 5282.55%

FCR4V6 FIRE RKAUX REL BFDS0A 4.50E08 2.47E08 82.24%

FCR4V7 FIRE RKAUX RPTSTB FDS0A 4.50E08 2.47E08 82.26%

FCR4N5 FIRE TurbineControlCabinet(RKEHC) FDS1 4.36E08 1.76E08 147.85%

FES2D8 FIRE 480VTRMR19P1 FDS3 4.10E08 3.32E08 23.46%

FDG101 FIRE 1DG 11VS F 22AFDS0 3.73E08 1.52E08 144.49%

FCR4B4 FIRE RKROD POS 1, 2, 3, 4FDS3 3.53E08 3.11E09 1034.80%

FNS1I1 FIRE TRANS1 3E FDS1 3.52E08 2.62E08 34.09%

FES2B7 FIRE 4KVS1DF (Sections1116) HEAF FDS2/4 3.39E08 2.86E08 18.66%

IE POW LOSSOFOFFSITEPOWEREXTREME LOSPE ER WEATHERRELATED 3.24E08 8.37E09 287.19%

FES2B3 FIRE 4KVS1DF (Sections310) HEAF FDS2/4 3.24E08 2.53E08 27.90%

FES2C5 FIRE TRANS1 9P FDS2 3.21E08 2.61E08 23.02%

FCR4B9 FIRE SER(RKSEQ) FDS3 3.12E08 1.28E08 142.84%

FCR4W7 FIRE 1CR 4125VDC DistPnlNo5FDS2 3.05E08 2.62E08 16.47%

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 19 of 24

B1R8HCL1 BV1REV8 Results Results Comparison IE %Changein Initiator Type Description CDF CDF CDF FCR4W4 FIRE 1CR 4125VDC DistPnlNo4FDS2 3.05E08 2.62E08 16.47%

FCR4D9 FIRE PNLREL 42 FDS3 2.96E08 3.60E09 722.49%

FCR4A4 FIRE RKCMPTERM3,4 FDS3 2.84E08 3.45E09 722.65%

FCV259 FIRE 1CV 2TS#9FDS3/6 2.76E08 8.18E09 237.21%

1CR 4DL 1RC100A ICCMonitorTrainA &

FCR4C4 FIRE BFDS3 2.73E08 3.32E09 722.84%

IE POW LOSSOFOFFSITEPOWERSWITCHYARD LOSPY ER CENTERED 2.73E08 6.20E09 340.00%

FES2A5 FIRE 4KVS1DF (Sections1116) FDS2 2.51E08 2.12E08 18.65%

FCR4P9 FIRE RK3OVAPROC 20 FDS3 2.51E08 1.08E08 132.17%

FCR4P4 FIRE RK1OVAPROC 10 FDS3 2.51E08 1.08E08 132.17%

FYAR64 FIRE Bin24 Scenario1 2.40E08 2.16E08 11.32%

FYAR89 FIRE Bin24 Scenario26 2.40E08 2.16E08 11.32%

FES2A1 FIRE 4KVS1DF (Sections310) FDS2 2.40E08 1.88E08 27.90%

FDG104 FIRE MCC120 FDS0 2.24E08 1.07E08 109.19%

FES255 FIRE MCC1E10 FDS0 2.22E08 1.69E08 31.68%

FCR4O3 FIRE 1CR 4PNLSHUTDOWNFDS4 2.19E08 1.13E08 93.34%

IE TRAN PLMFW S PARTIAL LOSSOFMAINFEEDWATER 2.18E08 1.78E08 22.15%

FCR4T7 FIRE TS#17Bins6&7 2.16E08 1.81E08 19.57%

IE POW DOX ER LOSSOFEMERGENCY125VDC ORANGE 2.16E08 7.66E09 182.39%

FCV258 FIRE TS#9FDS1/2/5 2.14E08 6.00E09 257.39%

FDG131 FIRE TS#7 2.10E08 1.01E08 107.41%

FCR4T6 FIRE TS#16Bins6&7 1.94E08 1.13E08 71.03%

IE TRAN RTRIP S REACTORTRIP 1.93E08 1.30E08 48.28%

FES208 FIRE BATCHG1 2 FDS1 1.83E08 1.41E08 29.16%

FRC164 FIRE 1RC 1BV LP 1 R1FDS1 1.80E08 1.43E08 25.70%

FRC169 FIRE 1RC 1BV LP 1 U1FDS1 1.80E08 1.43E08 25.64%

FRC168 FIRE 1RC 1BV LP 1 R7FDS1 1.80E08 1.43E08 25.64%

FRC167 FIRE 1RC 1BV LP 1 R4FDS1 1.80E08 1.43E08 25.64%

FRC166 FIRE 1RC 1BV LP 1 R3FDS1 1.80E08 1.43E08 25.64%

FRC165 FIRE 1RC 1BV LP 1 R2FDS1 1.80E08 1.43E08 25.64%

FRC163 FIRE 1RC 1BV LP 1 E3FDS1 1.80E08 1.43E08 25.64%

FDG112 FIRE EEC 1A FDS1 1.73E08 8.35E09 107.43%

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 20 of 24

B1R8HCL1 BV1REV8 Results Results Comparison IE %Changein Initiator Type Description CDF CDFCDF IE POW LOSSOFOFFSITEPOWERSEVEREWEATHER LOSPW ER RELATED 1.73E08 3.92E09 340.76%

FDG125 FIRE TS#1 1.71E08 8.24E09 107.43%

IE SGTRB LOCA STEAMGENERATORBTUBE RUPTURE 1.69E08 6.75E09 150.95%

FDG106 FIRE MCC1E7 FDS0 1.68E08 6.86E09 144.57%

FRC133 FIRE 1RC 1692S 1.67E08 1.33E08 25.69%

IE SGTRA LOCA STEAMGENERATORATUBERUPTURE 1.66E08 6.47E09 156.50%

IE SGTRC LOCA STEAMGENERATORCTUBE RUPTURE 1.66E08 6.47E09 156.51%

FES279 FIRE 480VUS1 9 FDS2 1.65E08 1.34E08 23.45%

FDG127 FIRE TS#3 1.63E08 7.77E09 109.22%

IE TRAN IMSIV S CLOSUREOFONEMSIV 1.54E08 3.74E09 312.30%

FRC132 FIRE 1RC 1692NE 1.53E08 1.22E08 25.68%

FES289 FIRE BKR480VUS 19P16 FDS2 1.51E08 1.28E08 18.63%

FCR4W9 FIRE 1CR 4RK AUX REL AFDS1A 1.51E08 6.96E09 117.31%

IE POW IRX ER LOSSOFVITALBUSI(RED) 1.49E08 2.07E09 620.98%

FCR4X1 FIRE 1CR 4RK AUX RPTST AFDS1A 1.49E08 6.71E09 121.79%

FRC158 FIRE 1RC 1767S 1.48E08 1.18E08 25.64%

IE TRAN TLMFW S TOTALLOSSOFMAINFEEDWATER 1.47E08 7.73E09 90.41%

FCR4F0 FIRE 1CR 4RK CMPDIN 6FDS3 1.42E08 6.12E09 132.29%

IE POW LOSPP ER LOSSOFOFFSITEPOWERPLANTCENTERED 1.28E08 2.89E09 341.40%

FMCR84 FIRE CMP 1.27E08 5.67E09 123.64%

FDG130 FIRE 1DG 1TS#6 1.24E08 5.08E09 144.62%

FES293 FIRE BKR480VUS 19P16 (HEAF)FDS2/4 1.13E08 9.51E09 18.62%

FES2A9 FIRE 4KVS1DF (Sections12) HEAFFDS2/4 1.13E08 9.50E09 18.64%

FDG123 FIRE Bin11 1.07E08 5.13E09 109.28%

IE TRAN LCV S LOSSOFCONDENSERVACUUM 1.01E08 3.86E09 161.99%

FCS1A3 FIRE TS#22DBIN6/7FDS1 1.00E08 5.51E09 82.32%

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 21 of 24

B1R8HCL1 BV1REV8 Results Results Comparison IE %Changein Initiator Type DescriptionCDF CDFCDF FCR452 FIRE CommunicationsBatteryCharger48AFDS2 9.53E09 6.51E09 46.57%

FMCR06 FIRE BB B 8.69E09 7.01E09 24.02%

FRC155 FIRE 767N 8.53E09 6.42E09 32.98%

FES297 FIRE 4KVS1DF (Sections12) FDS2 8.32E09 7.05E09 18.11%

IE TRAN EXFW S EXCESSIVEFEEDWATERFLOW 7.50E09 6.14E09 22.11%

FES209 FIRE BATCHG1 2 FDS2 7.22E09 6.09E09 18.63%

FES2C8 FIRE 1ES 2Bin 18 7.18E09 5.54E09 29.70%

FES248 FIRE INVVITBUS1 2FDS1 7.12E09 5.51E09 29.16%

FES241 FIRE DCSWBD2 FDS2 7.02E09 5.92E09 18.64%

FYAR26 FIRE 3YARD 1Bin 11 Scenario 261EMH4A 6.97E09 6.26E09 11.33%

IE TRAN LPRF S LOSSOFPRIMARYFLOW 6.78E09 1.45E09 367.63%

IE TRAN ISI S INADVERTANTSAFETYINJECTION INITIATION 6.31E09 1.53E09 313.45%

FCS1M4 FIRE 1CS 1BIN18 TS#248 5.80E09 5.23E09 10.86%

FCV257 FIRE TS#8 5.73E09 5.15E09 11.29%

FLOO USB4CA D INTERNALFLOOD:SB4CSE 2 5.03E09 4.57E09 10.02%

FLOO UTB4A D INTERNALFLOOD:TB4SE 1 4.87E09 2.38E09 104.31%

FYAR01 FIRE 3YARD 1Bin 11 Scenario 11EMH11A 4.14E09 3.72E09 11.33%

IE TRAN SLBC S STEAMLINEBREAKINCOMMONRHSLINE 2.53E09 2.28E09 11.04%

FLOO UTG1A D INTERNALFLOOD:TG1SE 2 2.52E09 1.23E09 104.44%

FCR4B3 FIRE RKROD POS 1, 2, 3, 4FDS2 2.52E09 1.93E09 30.14%

FCS1A9 FIRE 1CS 1TS#22F BIN6/7FDS1 2.40E09 2.09E09 14.80%

G01 SEIS SEISMICPGA(0.061 0.15 G) 2.31E09 2.05E09 12.91%

IE TRAN SLBD S STEAMLINEBREAKOUTSIDECONTAINMENT 1.97E09 1.07E09 85.05%

IE POW LB1A ER LOSSOFNORMAL4KV BUS1A 1.76E09 5.87E10 199.08%

FLOO UTB1A D INTERNALFLOOD:TB1SE 2 1.21E09 6.66E10 81.79%

PRA-BV1-23-003-R00 NOBP-CC-6002, Rev. 05, Att. 5 Page 22 of 24

B1R8HCL1 BV1REV8 Results Results Comparison IE %Changein Initiator Type DescriptionCDF CDFCDF IE TRAN CPEXC S COREPOWEREXCURSION 1.14E09 2.43E10 370.53%

FLOO UTB4B D INTERNALFLOOD:TB4SP 1 1.03E09 6.52E10 58.45%

IE TRAN IAX S LOSSOFSTATIONINSTRUMENT AIR 8.64E10 4.71E10 83.53%

IE POW LB1D ER LOSSOFNORMAL4KV BUS1D 6.61E10 5.97E10 10.71%

UPA1G FLOO D D INTERNALFLOOD:PA1GFWLP 2SA 5.70E10 4.40E10 29.50%

IE TRAN ICX S LOSSOFCONTAINMENTINSTRUMENT AIR 5.06E10 4.09E10 23.67%

FMS103 FIRE MCC118 FDS0/1 3.82E10 3.13E10 21.99%

IE TRAN MFWLB S MAINFEEDWATERLINEBREAK 3.73E10 1.65E10 125.88%

IE TRAN AMSIV S CLOSUREOFALL MSIV'S 3.53E10 2.38E10 48.68%

FMS101 FIRE MCC117 FDS0/1 3.49E10 2.86E10 22.01%

IE TRAN SLBI S STEAMLINE BREAKINSIDECONTAINMENT 3.45E10 2.67E10 29.14%

IE POW IBX ER LOSSOFVITALBUSIII(BLUE) 2.96E10 2.48E10 19.39%

IE TRAN MAINSTEAMRELIEFORSAFETYVALVE MSV S OPENING 1.02E10 2.33E11 335.90%

FLOO UVP1A D INTERNALFLOOD:VP1FWLL 1 REJ 7.24E11 5.60E11 29.32%

FES2D9 FIRE TRANS1 9P FDS3 3.43E11 8.38E12 309.80%

FNS177 FIRE 1NS14KVS 1A1 2344AFDS1 3.76E12 1.85E12 102.84%

v t e Letter Sequence Request
EPID:L-2023-LLA-0027, Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair (Approved, Closed)
Initiation Request Acceptance... Supplement Administration Questions Answers 4 March 2023 Results Approval Other:
MONTHYEAR2023-03-04 [Table View]

L-23-073, Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair

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SUBJECT:

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SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

SUMMARY

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4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

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L-23-073, Emergency License Amendment Request for Technical Specification 3.5.2 Regarding One-Time Action for Valve Leak Repair

Text

SUBJECT:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

Subject:

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