Revision as of 16:41, 29 October 2019

Text

Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2
	ML17275A520
Person / Time
Site:	FitzPatrick
Issue date:	10/02/2017
From:	Jim Barstow; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	JAFP-17-0093, TSTF-542, Rev 2
	Download: ML17275A520 (142)
	v • d • e

Exelon Generation 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.90 JAFP-17-0093 October 2, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 James A. FitzPatrick Nuclear Power Plant, Unit 1 Renewed Facility Operating License Nos. DPR-59 NRG Docket Nos. 50-333

Subject:

Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control," Revision 2 Pursuant to 1O CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGG) is submitting a request for an amendment to the Technical Specifications (TS) for James A. FitzPatrick Nuclear Power Plant (JAF), Unit 1.

The proposed changes replace existing Technical Specifications (TS) requirements related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel Water Inventory Control (RPV WIG) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel.

EGG has concluded that the proposed changes present no significant hazards consideration under the standards set forth in 1O CFR 50.92.

The proposed changes have been reviewed by the JAF On-Site Safety Review Committee in accordance with the requirements of the JAF Quality Assurance Program.

This amendment request contains no regulatory commitments.

Attachment 1 provides a description and assessment of the proposed changes.

Attachment 2 provides the existing TS pages marked up to show the proposed changes.

Attachment 3 provides the existing Bases pages marked up to show the proposed changes (information only).

U.S. Nuclear Regulatory Commission Application to Revise TS to Adopt TSTF-542 Docket Nos. 50-333 October 2, 2017 Page 2 EGC requests approval of the proposed amendment by August 16, 2018, in support of the Fall 2018 refueling outage. Once approved, the amendment shall be implemented prior to the next refueling outage.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), EGC is notifying the State of New York of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

If you have any questions or require additional information, please contact Christian Williams at (61 O) 765-5729.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 2nd day of October 2017.

Respectfully, J&:rato~

Director - Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Evaluation of Proposed Changes

2. Markup of Technical Specifications Pages

3. Markup of Technical Specifications Bases Pages (For Information Only) cc: USNRC Region I, Regional Administrator w/ attachments II USNRC Senior Resident Inspector, JAF USNRC Project Manager, JAF

ATTACHMENT 1 Description and Assessment of Proposed Changes James A. FitzPatrick Nuclear Power Plant Renewed Facility Operating License Nos. DPR-59 Docket Nos. 50-333

Subject:

Application to Revise Technical Specifications to Adopt TSTF-542, "Reactor Pressure Vessel Water Inventory Control," Revision 2

1.0 DESCRIPTION

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation 2.2 Variations

3.0 REGULATORY ANALYSIS

3.1 NO SIGNIFICANT HAZARDS CONSIDERATION

4.0 ENVIRONMENTAL CONSIDERATION

5.0 REFERENCES

Description and Assessment of Proposed Changes Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 1 of 7 Docket Nos. 50-333

1.0 DESCRIPTION

Exelon Generation Company, LLC (EGC), proposes changes to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. DPR-59 James A. FitzPatrick Nuclear Power Plant (JAF).

The proposed changes replace existing TS requirements related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel (RPV) Water Inventory Control to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel.

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation EGC has reviewed the safety evaluation provided to the Technical Specifications Task Force on December 20, 2016 (Reference 1), as well as the information provided in TSTF-542 (Reference 2). EGC has concluded that the justifications presented in TSTF-542 and the safety evaluation prepared by the Nuclear Regulatory Commission (NRC) staff are applicable to JAF and justify this amendment for the incorporation of the changes to the JAF TS.

The following JAF TS reference or are related to OPDRVs and are affected by the proposed changes:

3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation 3.3.6.1 Primary Containment Isolation Instrumentation 3.3.6.2 Secondary Containment Isolation Instrumentation 3.3.7.1 Control Room Emergency Ventilation Air Supply (CREVAS) System Instrumentation 3.5.2 ECCS - Shutdown 3.6.1.3 Primary Containment Isolation Valves (PCIVs) 3.6.4.1 Secondary Containment 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) 3.6.4.3 Standby Gas Treatment (SGT) System 3.7.3 Control Room Emergency Ventilation Air Supply (CREVAS) System 3.7.4 Control Room AC System 3.8.2 AC Sources - Shutdown 3.8.5 DC Sources - Shutdown 3.8.8 Distribution Systems - Shutdown 2.2 Variations EGC is proposing the following variations from the TS changes described in TSTF-542. These variations do not affect the applicability of TSTF-542 or the NRC staff's safety evaluation to the proposed license amendment.

2.2.1 The JAF TS utilize different numbering and titles than the Standard Technical Specifications (STS) on which TSTF-542 was based. The differences are administrative and do not affect the applicability of TSTF-542 to the JAF TS.

Description and Assessl')1ent of Proposed Changes Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 2 of 7 Docket Nos. 50-333

STS Table 3.3.5.1-1 Function 1.c, 2.c, and 2.d, in part, are titled "Reactor Steam Dome Pressure - Low. In JAF TS, Function 1.c, 2.c, and 2.d, in part, are titled "Reactor Pressure - Low.

STS Section 3.3.7.1 is titled "Main Control Room Environmental Control System Instrumentation". The equivalent section in JAF TS is titled "Control Room Emergency Ventilation Air Supply (CREVAS) System Instrumentation".

STS Section 3.7.4 is titled "Main Control Room Environmental Control System". The equivalent section in JAF TS is titled "Control Room Emergency Ventilation Air Supply (CREVAS) System".

The STS TS LCO 3.8.8, "Inverters - Shutdown", does not exist in the JAF TS. JAF TS LCO 3.8.8, "Distribution Systems - Shutdown" will be revised consistent with STS LCO 3.8.10 "Distribution Systems - Shutdown". This variation is editorial in nature.

The STS Section 3.8.1 O is titled "Distribution Systems - Shutdown". The equivalent JAF TS section is numbered 3.8.8.

2.2.2 The JAF TS contain a Surveillance Frequency Control Program. Therefore, the Surveillance Requirements (SR) Frequencies for Specifications 3.3.5.2 and 3.5.2 are "In accordance with the Surveillance Frequency Control Program". This variation is editorial in nature.

2.2.3 The JAF TS contain NOTES in SR 3.5.1.2 and SR 3.5.2.4 regarding alignment of the Low Pressure Coolant Injection mode that are the same as the NOTES in STS LCOs 3.5.1 and 3.5.2. JAF will relocate these NOTES from the SRs to the LCO section. This has no effect on the adoption of the TSTF-542 and increases consistency between the JAF TS and STS. This variation is editorial in nature.

2.2.4 The JAF TS Table 3.3.5.1-1 contains notes (a) thru (d). The STS Table 3.3.5.1-1 contains notes (a) thru (f). The deletion of note (a) from the STS and JAF TS will result in renaming the remaining notes (a) thru (c) for JAF as opposed to (a) thru (e) for the STS.

This variation is editorial in nature.

2.2.5 The JAF TS 3.3.7.1 APPLICABILITY includes "During operations with a potential for draining the reactor vessel." This applicability is captured in the STS as note (a) in Table 3.3.7.1-1. As this applicability is being removed, this variation is editorial in nature.

2.2.6 There are STS requirements on which TSTF-542 is based, related to "manual initiation,"

that do not appear in the JAF TS.

STS Table 3.3.5.1-1 contains Functions 1.e and 2.h, Manual Initiation, for CS and LPCI, respectfully. The "manual initiation" logic does not exist in the JAF design. These functions, as well as the related TSTF-542 surveillance requirements, SR 3.3.5.2.3 and SR 3.5.2.8, do not apply to JAF.

As an alternative, EGC proposes that TS 3.5.2, "Reactor Pressure Vessel (RPV)

Water Inventory Control," include an SR 3.5.2.8 to verify that the JAF required ECCS injection/spray subsystem can be manually operated through the manipulation of subsystem components from the Main Control Room. The manual operation of the required ECCS injection/spray subsystem for the control of reactor cavity or RPV inventory is a relatively simple evolution and involves the

Description and Assessment of Proposed Changes Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 3 of 7 Docket Nos. 50-333 manipulation of a small number of components. These subsystem alignments can be performed by licensed operators from the Main Control Room.

This alternative is justified by the fact that a draining event is a slow evolution when compared to a design basis loss of coolant accident, which is assumed to occur at full power, and thus, there is adequate time to take manual actions (i.e.,

hours versus minutes). Adequate time to take action is assured since the proposed TS 3.5.2, Condition E, prohibits plant conditions that result in drain times that are less than one hour. Therefore, there is sufficient time for the licensed operators to take manual action to stop an unanticipated draining event, and to manually start an ECCS injection/spray subsystem or the additional method of water injection.

Since the ECCS injection/spray subsystem can be placed in service using manual means in a short period of time (i.e., within the time frames assumed in the development of TSTF-542), using controls and indications that are readily available in the Main Control Room, manual operation of the required subsystem would be an equivalent alternative to system initiation via manual initiation logic.

Current SR 3.5.1.6 and SR 3.5.2.4 manually operate the ECCS injection/spray pumps to verify each required ECCS injection/spray pump develops the specified flow rate against a system head corresponding to the specified reactor pressure at a frequency specified by the lnservice Testing (IST} Program. The IST Program requires the ECCS injection/spray subsystems motor operated injection valves, minimum flow valves and test flow path valves (with the exception of the CS test flow path valves) be cycled to demonstrate operability and compliance with IST stroke time requirements at a frequency specified by the IST Program.

The CS test flow path valves are part of the IST Program but do not have stroke time requirements. The CS valves are cycled for position indication verification only.

The manual operation of the ECCS injection/spray subsystem to demonstrate operability required by the proposed SR 3.5.2.7 is equivalent to the testing that is presently required to be performed on the ECCS injection/spray subsystems.

This variation is technical in nature due to plant design differences and similar to variations described in the license amendment requests submitted for Dresden Nuclear Power Station, Edwin I. Hatch Nuclear Plant, and Cooper Nuclear Power Station.

2.2.7 The JAF TS contain requirements that differ from the STS on which TSTF-542 was based, but are encompassed in the TSTF-542 justification.

The JAF TS 3.3.6.1 Primary Containment Isolation Instrumentation Function 5e, Reactor Water Cleanup (RWCU) System Isolation, occurs at Reactor Water Vessel Level- Low, Level 3. The STS isolation signal occurs at Reactor Water Vessel Level - Low Low, Level 2. The intent of the TSTF is to change the applicability of this function, but does not change the setpoint or allowable value. This variation is technical however is consistent with the intent of TSTF 542.

JAF TS Table 3.3.5.1-1 contains Function 1.d, "Core Spray Pump Start -

Time Delay Relay", that does not appear in the STS table. The Function is required to be operable in Modes 1, 2, 3, 4 and 5. Modes 4 and 5 are

Description and Assessment of Proposed Changes Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 4 of 7 Docket Nos. 50-333 being deleted from this Function as this is related to automatic ECCS initiation. This is the same justification that is provided in the TSTF for STS Table 3.3.5.1-1, Function 2.f, "Low Pressure Coolant Injection Pump Start - Time Delay Relay". This variation is technical in nature and is justified by the discussion in Section 3.4.1 of the TSTF-542 justification.

JAF TS Table 3.3.5.1-1 contains Function 1.f, "Core Spray Pump Discharge Pressure - High (Bypass)", that does not appear in the STS table. The Function is required to be operable in Modes 1, 2, 3, 4 and 5.

The Function is required for protection of the low pressure ECCS pump from overheating when the associated injection valve is not fully open, similar to STS Function 1.d and 2.g, "Core Spray and Low Pressure Coolant Injection Pump Discharge Flow- Low (Bypass)". Modes 4 and 5 of JAF TS Function 1.f are being moved to the new TS 3.3.5.2 as Function 1.c. Justification for this move is consistent with the justification provided in TSTF-542 Section 3.4.2 for STS Functions 1.d and 2.g from Table 3.3.5.1-1.

2.2.8 JAF TS 3.6.1.3, PCIVs, Condition G applies to Modes 4 and 5. In the STS the corresponding applicable condition is Condition H, therefore the required action to Initiate Action to suspend OPDRVs applies to G.1 for JAF as opposed to H.1 for the STS. This variation is administrative in nature.

2.2.9 TSTF-542 Rev 2 Eliminated Required Action (RA) J.2 from TS 3.3.6.1. Due to editorial errors which occurred between Rev O and Rev 1, the associated BASES pages and justification for this deletion were removed from the TSTF. They have been included in this submittal as a variation from Rev 2 of TSTF-542 with the following justification:

JAF TS 3.3.6.1 RA J.2 is triggered by Function 6.b from Table 3.3.6.1 (Shutdown Cooling System Isolation: Reactor Vessel Water Level - Low, Level 3). The direction to initiate action to close the Residual Heat Removal (RHR) Shutdown Cooling (SOC) isolation valves in Mode 3 is in direct conflict with TS 3.4.7 (RHR Shutdown Cooling System - Hot Shutdown) which requires two RHR SOC subsystems to be operable, and if not, to take immediate action to restore an RHR SOC subsystem to operable status (RA A.1 ). Therefore, Required Action J.2 is being deleted.

Removing Required Action J.2 is also appropriate to protect plant safety.

As discussed in the Bases to Function 6.b, the Reactor Vessel Level -

Low, Level 3 Function associated with the RHR SOC System is not directly assumed in the safety analyses because a break of the RHR Shutdown Cooling System is bounded by breaks of the reactor recirculation system and main steam lines. Specifically, for the RHR SOC isolation valves to be open in Mode 3, reactor steam dome pressure would need to be below the RHR cut-in permissive pressure.

Should a LOCA occur inside Primary Containment, TS 3.5.1 explicitly credits the manual closing of the RHR SOC isolation valves and alignment of RHR in the LPCI mode. Similarly, if the break is on the RHR SOC system outside Primary Containment, credit can still be given for manual closing of the RHR SOC isolation valves and alignment of an intact LPCI loop. In either case, core uncovery would not result and

Description and Assessment of Proposed Changes Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 5 of 7 Docket Nos. 50-333 radiological consequences are bounded by the LOCA and MSLB accidents.

For these reasons, it is not critical to immediately initiate action to close the AHR SOC isolation valves (RA J.2) if Function 6.b.

2.2.10 TSTF-542 Rev 2 Enclosure 7, page 3.8.2-3 (SR 3.8.2-1 NOTE 2), contains an editorial error which is a reference to LCO 3.5.2, "ECCS - Shutdown". This is being revised in the JAF NOTE 2 of SR 3.8.2-1 to reference LCO 3.5.2, "Reactor Pressure Vessel (RPV)

Water Inventory Control." This variation is consistent with the intent of the TSTF and is administrative in nature.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Determination Exelon Generation Company, LLC (EGC), proposes changes to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. DPR-59 for the James A.

FitzPatrick Nuclear Power Station.

EGC requests adoption of TSTF-542 "Reactor Pressure Vessel Water Inventory Control," which is an approved change to the Standard Technical Specifications (STS), into the JAF TS. The proposed amendment replaces the existing requirements in the TS related to "operations with a potential for draining the reactor vessel" (OPDRVs) with new requirements on Reactor Pressure Vessel (RPV) Water Inventory Control (WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires reactor vessel water level to be greater than the top of active irradiated fuel.

EGC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 1O CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an accident previously evaluated, and therefore replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.

The proposed changes reduce the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times.

Description and Assessment of Proposed Changes Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 6 of 7 Docket Nos. 50-333 These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event.

The proposed changes reduce the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be Operable in certain conditions in Mode 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed.

The proposed changes reduce or eliminate some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event.

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

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

Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed changes will not alter the design function of the equipment involved. Under the proposed changes, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements.

The event of concern under the current requirements and the proposed changes are an unexpected draining event. The proposed changes do not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases.

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

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

Description and Assessment of Proposed Changes Attachment 1 Application to Revise TS to Adopt TSTF-542 Page 7 of 7 Docket Nos. 50-333 Response: No.

The proposed changes replace existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the T AF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 1O CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.0 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 1O 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 effluent 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 1O 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.

5.0 REFERENCES

1. TSTF-542, Revision 2, "Reactor Pressure Vessel Water Inventory Control," dated December 29, 2016.

2. Final Safety Evaluation of Technical Specifications Task Force Traveler TSTF-542, Revision 2, "Reactor Pressure Vessel Water Inventory Control" (TAC No. MF3487)

ADAMS Accession No. ML163438065

ATTACHMENT2 Markup of Technical Specifications Pages James A. FitzPatrick Nuclear Power Station Renewed Facility Operating License Nos. DPR-59 Docket Nos. 50-333 Revised Technical Specifications Pages Unit 1 TS Pages 3.3.5.3-3* 3.6.4.1-2 ii 3.3.5.3-4* 3.6.4.2-1 1.1-2a* 3.3.6.1-3 3.6.4.2-3 3.3.5.1-1 3.3.6.1-10 3.6.4.3-1 3.3.5.1-2 3.3.6.2-4 3.6.4.3-2 3.3.5.1-4 3.3.7.1-1 3.7.3-1 3.3.5.1-8 3.5.1-1 3.7.3-2 3.3.5.1-9 3.5.1-4 3.7.3-3 3.3.5.1-10 3.5.2-1 3.7.4-1 3.3.5.1-11 3.5.2-2 3.7.4-2 3.3.5.1-12 3.5.2-3 3.8.2-2 3.3.5.2-1

3.5.2-4 3.8.2-3 3.3.5.2-2* 3.5.2-5 3.8.2-4 3.3.5.2-3* 3.5.3-1 3.8.5-2 3.3.5.3-1* 3.6.1.3-6 3.8.8-2 3.3.5.3-2* 3.6.4.1-1

New TS Page

TABLE OF CONTENTS 1.0 USE AND APPLICATION 1.1 Definitions .......................................................................................... 1.1-1 1.2 Logical Connectors ............................................................................ 1.2-1 1.3 Completion Times .............................................................................. 1.3-1 1.4 Frequency .......................................................................................... 1.4-1 2.0 SAFETY LIMITS (SLs) 2.1 SLs ................................................................................................... 2.0-1 2.2 SL Violations ...................................................................................... 2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY .......... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ......................... 3.0-3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) ........................................................... 3.1.1-1 3.1.2 Reactivity Anomalies .......................................................................... 3.1.2-1 3.1.3 Control Rod OPERABILITY ............................................................... 3.1.3-1 3.1.4 Control Rod Scram Times .................................................................. 3.1.4-1 3.1.5 Control Rod Scram Accumulators ...................................................... 3.1.5-1 3.1.6 Rod Pattern Control ........................................................................... 3.1.6-1 3.1.7 Standby Liquid Control (SLC) System ................................................ 3.1.7-1 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves .................... 3.1.8-1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)3.2.1-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) .................................. 3.2.2-1 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) ................................... 3.2.3-1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation ............................ 3.3.1.1-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation................................... 3.3.1.2-1 3.3.2.1 Control Rod Block Instrumentation..................................................... 3.3.2.1-1 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation ................................................................... 3.3.2.2-1 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation ............................... 3.3.3.1-1 3.3.3.2 Remote Shutdown System ................................................................. 3.3.3.2-1 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation ................................... 3.3.4.1-1 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation ................ 3.3.5.1-1 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation.3.3.5.2-1 3.3.5.23 Reactor Core Isolation Cooling (RCIC) System Instrumentation .......................................................................... 3.3.5.23-1 3.3.6.1 Primary Containment Isolation Instrumentation .................................. 3.3.6.1-1 3.3.6.2 Secondary Containment Isolation Instrumentation ............................. 3.3.6.2-1 3.3.7.1 Control Room Emergency Ventilation Air Supply (CREVAS)

System Instrumentation ............................................................. 3.3.7.1-1 3.3.7.2 Condenser Air Removal Pump Isolation Instrumentation ................... 3.3.7.2-1 3.3.7.3 Emergency Service Water (ESW) System Instrumentation................ 3.3.7.3-1 (continued)

JAFNPP i Amendment 287XXX

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.8.1 Loss of Power (LOP) Instrumentation ........................................ 3.3.8.1-1 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring ........................................................................... 3.3.8.2-1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating ................................................... 3.4.1-1 3.4.2 Jet Pumps .................................................................................. 3.4.2-1 3.4.3 Safety/Relief Valves (S/RVs) ...................................................... 3.4.3-1 3.4.4 RCS Operational LEAKAGE....................................................... 3.4.4-1 3.4.5 RCS Leakage Detection Instrumentation ................................... 3.4.5-1 3.4.6 RCS Specific Activity .................................................................. 3.4.6-1 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown ..................................................... 3.4.7-1 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown ................................................... 3.4.8-1 3.4.9 RCS Pressure and Temperature (P/T) Limits ............................. 3.4.9-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating ..................................................................... 3.5.1-1 3.5.2 Reactor Pressure Vessel (RPV) Water Inventory ControlECCS -

Shutdown 3.5.2-1 3.5.3 RCIC System ............................................................................. 3.5.3-1 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary Containment ................................................................. 3.6.1.1-1 3.6.1.2 Primary Containment Air Locks .................................................. 3.6.1.2-1 3.6.1.3 Primary Containment Isolation Valves (PCIVs) .......................... 3.6.1.3-1 3.6.1.4 Drywell Pressure ........................................................................ 3.6.1.4-1 3.6.1.5 Drywell Air Temperature............................................................. 3.6.1.5-1 3.6.1.6 Reactor Building-to-Suppression Chamber Vacuum Breakers ... 3.6.1.6-1 3.6.1.7 Suppression Chamber-to-Drywell Vacuum Breakers.................. 3.6.1.7-1 3.6.1.8 Main Steam Leakage Collection (MSLC) System ....................... 3.6.1.8-1 3.6.1.9 Residual Heat Removal (RHR) Containment Spray System ....... 3.6.1.9-1 3.6.2.1 Suppression Pool Average Temperature.................................... 3.6.2.1-1 3.6.2.2 Suppression Pool Water Level ................................................... 3.6.2.2-1 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling ......... 3.6.2.3-1 3.6.2.4 Drywell-to-Suppression Chamber Differential Pressure.............. 3.6.2.4-1 3.6.3.1 Primary Containment Oxygen Concentration ............................. 3.6.3.1-1 3.6.3.2 Containment Atmosphere Dilution (CAD) System ...................... 3.6.3.2-1 3.6.4.1 Secondary Containment ............................................................. 3.6.4.1-1 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) ...................... 3.6.4.2-1 3.6.4.3 Standby Gas Treatment (SGT) System...................................... 3.6.4.3-1 3.7 PLANT SYSTEMS 3.7.1 Residual Heat Removal Service Water (RHRSW) System ......... 3.7.1-1 3.7.2 Emergency Service Water (ESW) System and Ultimate Heat Sink (UHS) ................................................................. 3.7.2-1 (continued)

JAFNPP ii Amendment 274XXX

1.1 Definitions (continued)

DRAIN TIME The DRAIN TIME is the time it would take for the water inventory in and above the Reactor Pressure Vessel (RPV) to drain to the top of the active fuel (TAF) seated in the RPV assuming:

a) The water inventory above the TAF is divided by the limiting drain rate:

b) The limiting drain rate is the larger of the drain rate through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e.g., seismic event, loss of normal power, single human error), for all penetration flow paths below the TAF except:

1. Penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed or otherwise secured in the closed position, blank flanges, or other devices that prevent flow of reactor coolant through the penetration flow paths;

2. Penetration flow paths capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation; or

3. Penetration flow paths with isolation devices that can be closed prior to the RPV water level being equal to the TAF by a dedicated operator trained in the task, who in continuous communication with the control room, is stationed at the controls, and is capable of closing the penetration flow path isolation device without offsite power.

c) The penetration flow paths required to be evaluated per paragraph b) are assumed to open instantaneously and are not subsequently isolated, and no water is assumed to be subsequently added to the RPV water inventory; d) No additional draining events occur; and e) Realistic cross-sectional areas and drain rates are used.

A bounding DRAIN TIME may be used in lieu of a calculated value.

(continued)

JAFNPP 1.1-2a Amendment XXX

ECCS Instrumentation 3.3.5.1 3.3 INSTRUMENTATION 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation LCO 3.3.5.1 The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.5.1-1.

ACTIONS

- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - - -- - - - - - - - - - - - - -

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5.1-1 for the channel.

B. As required by B.1 - - - - - -NOTES- - - - --

Required Action A.1 1. Only applicable and referenced in in MODES 1, 2, Table 3.3.5.1-1. and 3.

2. Only applicable for Functions 1.a, 1.b, 2.a, and 2.b.

Declare supported 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature(s) in both divisions.

AND (continued)

JAFNPP 3.3.5.1-1 Amendment 274XXX

ECCS Instrumentation 3.3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 - - - - - -NOTE - - - - - -

Only applicable for Functions 3.a and 3.b.

Declare High Pressure 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Coolant Injection discovery of (HPCI) System loss of HPCI inoperable. initiation capability AND B.3 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.

C. As required by C.1 - - - - - -NOTES- - - - - -

Required Action A.1 1. Only applicable and referenced in in MODES 1, 2, Table 3.3.5.1-1. and 3.

2. Only applicable for Functions 1.c, 1.d, 2.c, 2.d, and 2.f.

Declare supported 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature(s) in both divisions.

AND C.2 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

(continued)

JAFNPP 3.3.5.1-2 Amendment 274XXX

ECCS Instrumentation 3.3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. As required by E.1 - - - - - -NOTES- - - - - -

Required Action A.1 1. Only Applicable and referenced in in MODES 1, 2, Table 3.3.5.1-1. and 3.

2. Only applicable for Functions 1.e, 1.f, and 2.g.

Declare supported 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation Initiation capability capability for Is inoperable subsystems in both divisions.

AND E.2 Restore channel to 7 days OPERABLE status.

(continued)

JAFNPP 3.3.5.1-4 Amendment 274XXX

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 1 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

1. Core Spray System

a. Reactor Vessel Water 1, 2, 3, 4(ab) B SR 3.3.5.1.1 18 inches Level - Low Low Low 4(a), 5(a) SR 3.3.5.1.2 (Level 1) SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

b. Drywell 1, 2, 3 4(ab) B SR 3.3.5.1.1 2.7 psig Pressure - High SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

c. Reactor Pressure - Low 1, 2, 3 4 C SR 3.3.5.1.1 410 psig (Injection Permissive) SR 3.3.5.1.2 and SR 3.3.5.1.4 490 psig SR 3.3.5.1.5 SR 3.3.5.1.6 4(a), 5(a) 4 B SR 3.3.5.1.1 410 psig SR 3.3.5.1.2 and SR 3.3.5.1.4 490 psig SR 3.3.5.1.5 SR 3.3.5.1.6

d. Core Spray Pump 1, 2, 3, 1 per pump C SR 3.3.5.1.5 12.34 Start - Time Delay 4(a), 5(a) SR 3.3.5.1.6 seconds Relay

e. Core Spray Pump 1, 2, 3, 1 per pump E SR 3.3.5.1.5 510 gpm Discharge Flow - Low 4(a), 5(a) SR 3.3.5.1.6 and (Bypass) 980 gpm

f. Core Spray Pump 1, 2, 3, 1 per pump E SR 3.3.5.1.3 90 psig and Discharge 4(a), 5(a) SR 3.3.5.1.6 110 psig Pressure - High (Bypass)

2. Low Pressure Coolant Injection (LPCI) System

a. Reactor Vessel Water 1, 2, 3, 4(ab) B SR 3.3.5.1.1 18 inches Level - Low Low Low 4(a), 5(a) SR 3.3.5.1.2 (Level 1) SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6 (continued)

(a) When associated ECCS subsystem(s) are required to be OPERABLE per LCO 3.5.2, ECCS - Shutdown.

(ba) Also required to initiate the associated emergency diesel generator subsystem.

JAFNPP 3.3.5.1-8 Amendment 274XXX

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

2. LPCI System (continued)

b. Drywell 1, 2, 3 4(ab) B SR 3.3.5.1.1 2.7 psig Pressure - High SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

c. Reactor Pressure - Low 1, 2, 3 4 C SR 3.3.5.1.1 410 psig (Injection Permissive) SR 3.3.5.1.2 and SR 3.3.5.1.4 490 psig SR 3.3.5.1.5 SR 3.3.5.1.6 4(a), 5(a) 4 B SR 3.3.5.1.1 410 psig SR 3.3.5.1.2 and SR 3.3.5.1.4 490 psig SR 3.3.5.1.5 SR 3.3.5.1.6

d. Reactor Pressure - Low 1(bc), 2(bc), 4 C SR 3.3.5.1.1 295 psig (Recirculation 3(bc) SR 3.3.5.1.2 Discharge Valve SR 3.3.5.1.4 Permissive) SR 3.3.5.1.5 SR 3.3.5.1.6

e. Reactor Vessel Shroud 1, 2, 3 2 B SR 3.3.5.1.1 1.0 inches Level (Level 0) SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

f. Low Pressure Coolant 1, 2, 3, 1 per pump C SR 3.3.5.1.5 Injection Pump 4(a), 5(a) SR 3.3.5.1.6 Start - Time Delay Relay Pumps A, D 1.51 seconds Pumps B, C 6.73 seconds (continued)

(a) When associated ECCS subsystem(s) are required to be OPERABLE per LCO 3.5.2.

(ba) Also required to initiate the associated emergency diesel generator subsystem.

(cb) With associated recirculation pump discharge valve open.

JAFNPP 3.3.5.1-9 Amendment 274XXX

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

2. LPCI System (continued)

g. Low Pressure 1, 2, 3, 1 per E SR 3.3.5.1.5 1040 gpm Coolant Injection Pump 4(a), 5(a) subsystem SR 3.3.5.1.6 and Discharge Flow - Low 1665 gpm (Bypass)

h. Containment 1, 2, 3 4 B SR 3.3.5.1.3 1 psig and Pressure - High SR 3.3.5.1.6 2.7 psig

3. High Pressure Coolant Injection (HPCI) System

a. Reactor Vessel Water 1, 4 B SR 3.3.5.1.1 126.5 Level - Low Low 2(cd), 3(cd) SR 3.3.5.1.2 inches (Level 2) SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

b. Drywell 1, 4 B SR 3.3.5.1.1 2.7 psig Pressure - High 2(cd), 3(cd) SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

c. Reactor Vessel Water 1, 2 C SR 3.3.5.1.1 222.5 Level - High (Level 8) 2(cd), 3(cd) SR 3.3.5.1.2 inches SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

d. Condensate Storage 1, 4 D SR 3.3.5.1.3 59.5 inches Tank Level - Low 2(cd), 3(cd) SR 3.3.5.1.6

e. Suppression Pool Water 1, 2 D SR 3.3.5.1.3 14.5 ft Level - High 2(cd), 3(cd) SR 3.3.5.1.6

f. High Pressure Coolant 1, 1 E SR 3.3.5.1.5 475 gpm Injection Pump 2(cd), 3(cd) SR 3.3.5.1.6 and Discharge Flow - Low 800 gpm (Bypass)

g. High Pressure Coolant 1, 1 E SR 3.3.5.1.3 25 psig Injection Pump 2(cd), 3(cd) SR 3.3.5.1.6 and Discharge 80 psig Pressure - High (Bypass)

(continued)

(a) When associated ECCS subsystem(s) are required to be OPERABLE per LCO 3.5.2.

(dc) With reactor steam dose pressure > 150 psig.

JAFNPP 3.3.5.1-10 Amendment 274XXX

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 4 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

4. Automatic Depressurization System (ADS) Trip System A

a. Reactor Vessel Water 1, 2 F SR 3.3.5.1.1 18 inches Level - Low Low Low 2(cd), 3(cd) SR 3.3.5.1.2 (Level 1) SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

b. Automatic 1, 1 G SR 3.3.5.1.5 134 seconds Depressurization 2(cd), 3(cd) SR 3.3.5.1.6 System Initiation Timer

c. Reactor Vessel Water 1, 1 F SR 3.3.5.1.1 177 inches Level - Low (Level 3) 2(cd), 3(cd) SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

d. Core Spray Pump 1, 2 G SR 3.3.5.1.3 90 psig and Discharge 2(cd), 3(cd) SR 3.3.5.1.6 110 psig Pressure - High

e. Low Pressure Coolant 1, 4 G SR 3.3.5.1.3 105 psig Injection Pump 2(cd), 3(cd) SR 3.3.5.1.6 and Discharge 145 psig Pressure - High

5. ADS Trip System B

a. Reactor Vessel Water 1, 2 F SR 3.3.5.1.1 18 inches Level - Low Low Low 2(cd), 3(cd) SR 3.3.5.1.2 (Level 1) SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

b. Automatic 1, 1 G SR 3.3.5.1.5 134 seconds Depressurization 2(cd), 3(cd) SR 3.3.5.1.6 System Initiation Timer (continued)

(dc) With reactor steam dome pressure > 150 psig JAFNPP 3.3.5.1-11 Amendment 274XXX

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 5 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

5. ADS Trip System B (continued)

c. Reactor Vessel Water 1, 1 F SR 3.3.5.1.1 177 inches Level - Low (Level 3) 2(cd), 3(cd) SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

d. Core Spray Pump 1, 2 G SR 3.3.5.1.3 90 psig Discharge 2(cd), 3(cd) SR 3.3.5.1.6 and Pressure - High 110 psig

e. Low Pressure Coolant 1, 4 G SR 3.3.5.1.3 105 psig Injection Pump 2(cd), 3(cd) SR 3.3.5.1.6 and Discharge 145 psig Pressure - High (dc) With reactor steam dose pressure > 150 psig.

JAFNPP 3.3.5.1-12 Amendment 274XXX

RPV Water Inventory Control Instrumentation 3.3.5.2 3.3 INSTRUMENTATION 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation LCO 3.3.5.2 The Reactor Pressure Vessel (RPV) Water Inventory Control instrumentation for each Function in Table 3.3.5.2-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.5.2-1.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5.2-1 for the channel.

B. As required by Required B.1 Declare associated Immediately Action A.1 and penetration flow path(s) referenced in incapable of automatic Table 3.3.5.2-1. isolation.

AND B.2 Calculate DRAIN TIME. Immediately C. As required by Required C.1 Place channel in trip. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Action A.1 and referenced in Table 3.3.5.2-1.

D. As required by Required D.1 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Action A.1 and OPERABLE status.

referenced in Table 3.3.5.2-1.

JAFNPP 3.3.5.2-1 Amendment XXX

RPV Water Inventory Control Instrumentation 3.3.5.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Declare associated low Immediately associated Completion pressure ECCS Time of Condition C or D injection/spray subsystem not met. inoperable.

SURVEILLANCE REQUIREMENTS

NOTE-----------------------------------------------------------

Refer to Table 3.3.5.2-1 to determine which SRs apply for each ECCS Function.

SURVEILLANCE FREQUENCY SR 3.3.5.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.2.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.5.2.3 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program JAFNPP 3.3.5.2-2 Amendment XXX

RPV Water Inventory Control Instrumentation 3.3.5.2 Table 3.3.5.2-1 (page 1 of 1)

Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

1. Core Spray

a. Reactor 4, 5 4 C SR 3.3.5.2.1 < 490 psig Pressure - SR 3.3.5.2.2 Low (Injection Permissive)

b. Core Spray 4, 5 1 per pump D SR 3.3.5.2.1 > 510 gpm Pump (a) SR 3.3.5.2.2 and Discharge < 980 gpm Flow - Low (Bypass)

c. Core Spray 4, 5 1 per pump D SR 3.3.5.2.1 > 90 psig Pump (a) SR 3.3.5.2.2 and Discharge < 110 psig Pressure -

High (Bypass)

2. Low Pressure Coolant Injection (LPCI) System

a. Reactor 4, 5 4 C SR 3.3.5.2.1 < 490 psig Pressure - SR 3.3.5.2.2 Low (Injection Permissive)

b. Low Pressure 4, 5 1 per pump D SR 3.3.5.2.1 > 1040 gpm Coolant (a) SR 3.3.5.2.2 and Injection < 1665 gpm Pump Discharge Flow - Low (Bypass)

3. RHR System Isolation

a. Reactor (b) 2 in one trip B SR 3.3.5.2.1 > 177 inches Vessel Water system SR 3.3.5.2.2 Level - Low, Level 3

4. Reactor Water Cleanup (RWCU) System Isolation

a. Reactor (b) 2 in one trip B SR 3.3.5.2.1 > 177 inches Vessel Water system SR 3.3.5.2.2 Level - Low Level 3 (a) Associated with an ECCS subsystem required to be OPERABLE by LCO 3.5.2, Reactor Pressure Vessel (RPV) Water Inventory Control.

(b) When automatic isolation of the associated penetration flow path(s) is credited in calculating DRAIN TIME.

JAFNPP 3.3.5.2-3 Amendment XXX

RCIC System Instrumentation 3.3.5.32 3.3 INSTRUMENTATION 3.3.5.32 Reactor Core Isolation Cooling (RCIC) System Instrumentation LCO 3.3.5.32 The RCIC System instrumentation for each Function in Table 3.3.5.32-1 shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig ACTIONS

- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - - -- - - - - - - - - - - - - -

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5.32-1 for the channel.

B. As required by B.1 Declare RCIC System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Required Action A.1 inoperable. discovery of and referenced in loss of RCIC Table 3.3.5.32-1. initiation capability AND B.2 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.

C. As required by C.1 Restore channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Required Action A.1 OPERABLE status.

and referenced in Table 3.3.5.32-1.

(continued)

JAFNPP 3.3.5.32-1 Amendment 274XXX

RCIC System Instrumentation 3.3.5.32 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by D.1 - - - - - -NOTE- - - - - - -

Required Action A.1 Only applicable if and referenced in RCIC pump suction is Table 3.3.5.32-1. not aligned to the suppression pool.

Declare (RCIC) System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from inoperable. discovery of loss of automatic RCIC initiation capability AND D.2.1 Place channel in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months trip.

OR D.2.2 Align RCIC pump 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months suction to the suppression pool.

E. Required Action and E.1 Declare RCIC System Immediately associated Completion inoperable.

Time of Condition B, C, or D not met.

JAFNPP 3.3.5.32-2 Amendment 274XXX

RCIC System Instrumentation 3.3.5.32 SURVEILLANCE REQUIREMENTS

- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -NOTES- - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

1. Refer to Table 3.3.5.32-1 to determine which SRs apply for each RCIC Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 2 and 4; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 1 and 3 provided the associated Function maintains RCIC initiation capability.

SURVELLANCE FREQUENCY SR 3.3.5.32.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.32.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.5.32.3 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.5.32.4 Calibrate the trip units. In accordance with the Surveillance Frequency Control Program SR 3.3.5.32.5 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.5.32.6 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program JAFNPP 3.3.5.32-3 Amendment 301XXX

RCIC System Instrumentation 3.3.5.32 Table 3.3.5.32-1 (page 1 of 1)

Reactor Core Isolation Cooling System Instrumentation CONDITIONS REQUIRED REFERENCED CHANNELS FROM REQUIRED SURVEILLANCE ALLOWABLE FUNCTION PER FUNCTION ACTION A.1 REQUIREMENTS VALUE

1. Reactor Vessel Water 4 B SR 3.3.5.32.1 126.5 inches Level - Low Low (Level 2) SR 3.3.5.32.2 SR 3.3.5.32.4 SR 3.3.5.32.5 SR 3.3.5.32.6

2. Reactor Vessel Water 2 C SR 3.3.5.32.1 222.5 inches Level - High (Level 8) SR 3.3.5.32.2 SR 3.3.5.32.4 SR 3.3.5.32.5 SR 3.3.5.32.6

3. Condensate Storage Tank 4 D SR 3.3.5.32.3 59.5 inches Level - Low SR 3.3.5.32.6

4. Manual Initiation 1 C SR 3.3.5.32.6 NA JAFNPP 3.3.5.32-4 Amendment 274XXX

Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME H. Required Action and H.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition F or AND G not met.

H.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR As required by Required Action C.1 and referenced in Table 3.3.6.1-1.

I. As required by I.1 Declare associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Required Action C.1 standby liquid and referenced in control subsystem Table 3.3.6.1-1. (SLC) inoperable.

OR I.2 Isolate the Reactor 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Water Cleanup System.

J. As required by J.1 Initiate action to Immediately Required Action C.1 restore channel to and referenced in OPERABLE status.

Table 3.3.6.1-1.

OR J.2 Initiate action to Immediately isolate the Residual Heat Removal (RHR)

Shutdown Cooling System.

JAFNPP 3.3.6.1-3 Amendment 274XXX

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 5 of 6)

Primary Containment Isolation Instrumentation APPLICABLE REQUIRED CONDITIONS SURVEILLANCE ALLOWABLE MODES OR CHANNELS REFERENCED REQUIREMENTS VALUE OTHER PER TRIP FROM SPECIFIED SYSTEM REQUIRED FUNCTION CONDITIONS ACTION C.1

5. Reactor Water Cleanup (RWCU)

System Isolation

a. RWCU Suction Line 1,2,3 1 F SR 3.3.6.1.3 144°F Penetration Area SR 3.3.6.1.7 Temperature High

b. RWCU Pump Area 1,2,3 1 per F SR 3.3.6.1.3 165°F for Pump Temperature High room SR 3.3.6.1.7 Room A and 175°F for Pump Room B

c. RWCU Heat 1,2,3 1 F SR 3.3.6.1.3 155°F Exchanger Room Area SR 3.3.6.1.7 Temperature High

d. SLC System Initiation 1,2 2(d) I SR 3.3.6.1.7 NA

e. Reactor Vessel Water Level 1,2,3 2 F SR 3.3.6.1.1 177 inches Low (Level 3) SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7

f. Drywell Pressure High 1,2,3 2 F SR 3.3.6.1.1 2.7 psig SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7

6. Shutdown Cooling System Isolation

a. Reactor Pressure High 1,2,3 1 F SR 3.3.6.1.1 74 psig SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7

b. Reactor Vessel Water Level 3,4,5 2(e) J SR 3.3.6.1.1 177 inches Low (Level 3) SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7 (continued)

(d) SLC System Initiation only inputs into one of the two trip systems and only isolates one valve in the RWCU suction and return line.

(e) Only one trip system required in MODES 4 and 5 when RHR Shutdown Cooling System integrity maintained.

JAFNPP 3.3.6.1-10 Amendment 298XXX

Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)

Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water 1, 2, 3, 2 SR 3.3.6.2.1 177 inches Level - Low (Level 3) (a) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.5 SR 3.3.6.2.6

2. Drywell Pressure - High 1, 2, 3 2 SR 3.3.6.2.1 2.7 psig SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.5 SR 3.3.6.2.6

3. Reactor Building Exhaust 1, 2, 3, 1 SR 3.3.6.2.1 24,800 cpm Radiation - High (a), (b) SR 3.3.6.2.3 SR 3.3.6.2.6

4. Refueling Floor Exhaust 1, 2, 3, 1 SR 3.3.6.2.1 24,800 cpm Radiation - High (a), (b) SR 3.3.6.2.3 SR 3.3.6.2.6 (a) During operation with a potential for draining the reactor vessel.

(b)(a) During movement of recently irradiated fuel assemblies in secondary containment.

JAFNPP 3.3.6.2-4 Amendment 276XXX

`

CREVAS System Instrumentation 3.3.7.1 3.3 INSTRUMENTATION 3.3.7.1 Control Room Emergency Ventilation Air Supply (CREVAS)

System Instrumentation LCO 3.3.7.1 The Control Room Air Inlet Radiation High channel shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Channel inoperable. A.1 Place the CREVAS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months System in the isolate mode of operation.

OR A.2 Declare both CREVAS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months subsystems inoperable.

JAFNPP 3.3.7.1-1 Amendment 276XXX

ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS-Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

- - - - - - - - - - - - - - - -- - - - - - - - - -NOTE - - - - - - - - -- - - - - - -- - - - - - -

Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) cut in permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.

- - - - - - - - - - - - - - - - - -- - - - - - - - - - - - -- - - - - - - - - -- - - -- - - - - -- - - -

APPLICABILITY: MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

ACTIONS

NOTE-------------------------------------------------------------

LCO 3.0.4.b is not applicable to HPCI.

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.1 Restore low pressure 7 days injection/spray ECCS injection/spray subsystem inoperable. subsystem(s) to OR OPERABLE status.

One low pressure coolant injection (LPCI) pump in both LPCI subsystems inoperable.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in Mode 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

JAFNPP 3.5.1-1 Amendment 284XXX

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.2 - - - - - - - - - - - - - - - - - - - - - -NOTE - - - - - - - - - - - - -

Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) cut in permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.

- - - - - - - - - - - - - - - - - -- - - - - - - - - - - - -- - - - - - - - - -

Verify each ECCS injection/spray subsystem manual, In accordance with power operated, and automatic valve in the flow path, the Surveillance that is not locked, sealed, or otherwise secured in Frequency Control position, is in the correct position. Program SR 3.5.1.3 Verify ADS pneumatic supply header pressure is In accordance with

> 95 psig. the Surveillance Frequency Control Program SR 3.5.1.4 Verify the RHR System cross tie valves are closed and In accordance with power is removed from the electrical valve operator. the Surveillance Frequency Control Program SR 3.5.1.5 Cycle open and closed each LPCI motor operated valve In accordance with independent power supply battery charger AC input the Surveillance breaker and verify each LPCI inverter output voltage is Frequency Control

> 576 V and < 624 V while supplying the respective Program bus.

(continued)

JAFNPP 3.5.1-4 Amendment 301XXX

ECCS -- ShutdownRPV Water Inventory Control 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.2 ECCS -- ShutdownReactor Pressure Vessel (RPV) Water Inventory Control LCO 3.5.2 DRAIN TIME of RPV water inventory to the top of active fuel (TAF) shall be > 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months AND One Two low pressure ECCS injection/spray subsystems shall be OPERABLE.

- - - - - - - - - - - - - - - - - - - - - -NOTE - - - - - - - - - - - - -

A Low Pressure Coolant Injection (LPCI) subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

- - - - - - - - - - - - - - - - - -- - - - - - - - - - - - -- - - - - - - - - -

APPLICABILITY: MODES 4, and 5 MODE 5, except with the spent fuel storage pool gates removed and water level 22 ft 2 inches over the top of the reactor pressure vessel flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One rRequired low A.1 Restore required low 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months pressure ECCS pressure ECCS injection/spray injection/spray subsystem inoperable. subsystem to OPERABLE status.

B. Required Action and B.1 Initiate action to Immediately associated Completion establish a method of Time of Condition A water injection capable not met. of operating without offsite electrical power.suspend operations with a potential for draining the reactor vessel (OPDRVs).

JAFNPP 3.5.2-1 Amendment 274XXX

ECCS -- ShutdownRPV Water Inventory Control 3.5.2 C. DRAIN TIME < 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C.1 Verify secondary Immediately4 hours and > 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Two containment boundary is required low capable of being pressure ECCS established in less than injection/spray the DRAIN TIMEInitiate C. subsystems inoperable. action to suspend OPDRVs.

AND C.2 Verify each secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months containment penetration flow path is capable of being isolated in less than the DRAIN TIME.Restore one required low pressure ECCS injection/spray subsystem to OPERBLE status.

AND C.3 Verify one standby gas 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months treatment subsystem is capable of operation in less than the DRAIN TIME (continued)

JAFNPP 3.5.2-2 Amendment 274XXX

ECCS -- ShutdownRPV Water Inventory Control 3.5.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action C.2 and D.1 ---------NOTE------------ Immediately associated Completion Required ECCS Time not met.DRAIN TIME injection/spray subsystem

< 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or additional method of water injection shall be capable of operating without offsite electrical power.

Initiate action to restore secondary containment to OPERABLE status.establish an additional method of water injection with water sources capable of maintaining RPV water level

> TAF for > 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

AND D.2 Initiate action to establish Immediately secondary containment boundary restore one standby gas treatment subsystem to OPERABLE status.

AND D.3 Initiate action to isolate Immediately each secondary containment penetration flow path or verify it can be manually isolated from the control room.restore Isolation capability in each required secondary containment penetration flow path not isolated.

AND D.4 Initiate action to verify one Immediately standby gas treatment JAFNPP 3.5.2-2 Amendment XXX

ECCS -- ShutdownRPV Water Inventory Control 3.5.2 subsystem is capable of being placed in operation E. Required Action and E.1 Initiate action to restore Immediately associated Completion DRAIN TIME to > 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Time of Condition C or D not met.

OR DRAIN TIME < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SURVEILLANCE REQUIREMENTS SURVELLANCE FREQUENCY SR 3.5.2.1 Verify DRAIN TIME > 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months In accordance with the Surveillance Frequency Control Program SR 3.5.2.21 Verify, for each a required low pressure coolant injection In accordance with (LPCI) subsystem, the suppression pool water level is the Surveillance 10.33 ft. Frequency Control Program (continued)

JAFNPP 3.5.2-3 Amendment XXX

ECCS -- ShutdownRPV Water Inventory Control 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.32 Verify, for each a required cCore sSpray (CS) subsystem, In accordance the: with the Surveillance

a. Suppression pool water level is 10.33 ft; or Frequency Control Program

b. -------------------NOTE------------------------------

Only one required CS subsystem may take credit for this option during OPDRVs.

The water level in each condensate storage Tank is > 324 inches SR 3.5.2.43 Verify, for eachthe required ECCS injection/spray In accordance subsystem, the piping is filled with water from the with the pump discharge valve to the injection valve. Surveillance Frequency Control Program SR 3.5.2.54 - - - - - - - - - - - - - - - - - - - - - -NOTE - - - - - - - - - - - - - In accordance One LPCI subsystem may be considered with the OPERABLE during alignment and operation for Surveillance decay heat removal if capable of being manually Frequency Control realigned and not otherwise inoperable. Program

- - - - - - - - - - - - - - - - - -- - - - - - - - - - - - -- - - - - - - - - -

Verify, for the each required ECCS injection/spray subsystem, each 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.

(continued)

JAFNPP 3.5.2-4 Amendment XXX

ECCS -- ShutdownRPV Water Inventory Control 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVELLANCE FREQUENCY SR 3.5.2.65 Operate the required ECCS injection/spray subsystem In accordance with through the recirculation line for > 10 minutesVerify the Inservice Testing each required ECCS pump develops the ProgramSurveillance specified flow rate against a system head Frequency Control Program corresponding to the specified reactor pressure above primary containment pressure.

SYSTEM HEAD CORRESPONDING TO A REACTOR PRESSURE NO. ABOVE PRIMARY OF CONTAINMENT SYSTEM FLOW RATE PUMPS PRESSURE OF CS 4265 gpm 1 113 psi LPCI 7700 gpm 1 20 psi SR 3.5.2.7 Verify each valve credited for automatically In accordance with isolating a penetration flow path actuates to the the Surveillance isolation position on an actual or simulated isolation Frequency Control signal. Program SR 3.5.2.86 - - - - - - - - - - - - - - -NOTE - - - - - - - - - - - - - - - -

Vessel injection/spray may be excluded.

Verify each the required ECCS injection/spray subsystem In accordance with can be manually operated. actuates on an actual or the Surveillance simulated automatic Frequency Control initiation signal. Program JAFNPP 3.5.2-5 Amendment XXX

RCIC SYSTEM 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig.

ACTIONS

NOTE--------------------------------

LCO 3.0.4.b is not applicable to RCIC.

CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System inoperable. A.1 Verify by administrative Immediately means High Pressure Coolant Injection System is OPERABLE.

AND A.2 Restore RCIC System to 14 days OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND B.2 Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months dome pressure to 150 psig.

JAFNPP 3.5.3-1 Amendment 284XXX

PCIVs 3.6.1.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and F.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, AND B, C, D, or E not met in MODE 1, 2, or 3. F.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months G. Required Action and G.1 Initiate action to Immediately associated Completion suspend operations Time of Condition A or with a potential for B not met for PCIV(s) draining the reactor required to be vessel.

OPERABLE during Mode 4 or 5. OR G.12 Initiate action to Immediately restore valve(s) to OPERABLE status.

JAFNPP 3.6.1.3-6 Amendment 274XXX

Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment A.1 Restore secondary 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months inoperable in MODE 1, containment to 2, or 3. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Secondary containment C.1 ------------NOTE------------

inoperable during LCO 3.0.3 is not movement of recently applicable.

irradiated fuel ---------------------------------

assemblies in the secondary containment Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in the secondary containment.

AND (continued)

JAFNPP 3.6.4.1-1 Amendment 276XXX

Secondary Containment 3.6.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 Initiate action to suspend Immediately OPDVRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 0.25 inch In accordance with of vacuum water gauge. the Surveillance Frequency Control Program SR 3.6.4.1.2 Verify all secondary containment equipment hatches In accordance with are closed and sealed. the Surveillance Frequency Control Program SR 3.6.4.1.3 Verify one secondary containment access door in each In accordance with access opening is closed. the Surveillance Frequency Control Program SR 3.6.4.1.4 Verify the secondary containment can be maintained In accordance with 0.25 inch of vacuum water gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using the Surveillance one SGT subsystem at a flow rate 6000 cfm. Frequency Control Program JAFNPP 3.6.4.1-2 Amendment 301XXX

SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

LCO 3.6.4.2 Each SCIV shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS

NOTES----------------------------------------------------

1. Penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

2. Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months penetration flow paths penetration flow path with one SCIV by use of at least inoperable. one closed and de-activated automatic valve, closed manual valve, or blind flange.

AND (continued)

JAFNPP 3.6.4.2-1 Amendment 276XXX

SCIVs 3.6.4.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 --------NOTE---------

associated Completion LCO 3.0.3 is not Time of Condition A applicable.

or B not met during ---------------------

movement of recently irradiated fuel Suspend movement of Immediately assemblies in the recently irradiated secondary containment fuel assemblies in or during OPDRVs. the secondary containment.

AND Immediately D.2 Initiate action to suspend OPDRVs.

JAFNPP 3.6.4.2-3 Amendment 276XXX

SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT subsystem 7 days inoperable. to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Required Action and ----------------NOTE----------------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A ----------------------------------------

not met during C.1 Place OPERABLE SGT Immediately movement of recently subsystem in irradiated fuel operation.

assemblies in the secondary containment OR or during OPDRVs.

(continued)

JAFNPP 3.6.4.3-1 Amendment 276XXX

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2.1 Suspend movement of Immediately recently irradiated fuel assemblies in secondary containment.

AND Immediately C.2.2 Initiate action to suspend OPDRVs.

D. Two SGT subsystems D.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2, or 3.

E. Two SGT subsystems E.1 --------NOTE---------

inoperable during LCO 3.0.3 is not movement of recently applicable.

irradiated fuel -------------------------

assemblies in the Suspend movement of Immediately secondary containment recently irradiated or during OPDRVs. fuel assemblies in secondary containment.

AND E.2 Initiate action to Immediately suspend OPDRVs.

JAFNPP 3.6.4.3-2 Amendment 276XXX

CREVAS System 3.7.3 3.7 PLANT SYSTEMS 3.7.3 Control Room Emergency Ventilation Air Supply (CREVAS) System LCO 3.7.3 Two CREVAS subsystems shall be OPERABLE.

NOTE --------------------------------------------

The control room envelope (CRE) boundary may be opened intermittently under administrative control.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREVAS subsystem A.1 Restore CREVAS 7 days inoperable for reasons subsystem to OPERABLE other than Condition B. status.

B. One or more CREVAS B.1 Initiate action to implement Immediately subsystems inoperable due mitigating actions.

to inoperable CRE boundary in MODE 1, 2, or 3. AND B.2 Verify mitigating actions 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits.

AND B.3 Restore CRE boundary to 90 days OPERABLE status.

(continued)

JAFNPP 3.7.3-1 Amendment 289XXX

CREVAS System 3.7.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A or B not met AND in MODE 1, 2, or 3.

C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. Required Action and -------------- NOTE --------------

associated Completion Time LCO 3.0.3 is not applicable.

of Condition A not met --------------------------------------

during movement of recently irradiated fuel D.1 Place OPERABLE CREVAS Immediately assemblies in the secondary subsystem in isolate mode.

containment or during OPDRVs. OR D.2.1 Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.

AND D.2.2 Initiate action to suspend Immediately OPDRVs.

E. Two CREVAS subsystems E.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2, or 3 for reasons other than Condition B.

(continued)

JAFNPP 3.7.3-2 Amendment 289XXX

CREVAS System 3.7.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. Two CREVAS subsystems --------------- NOTE ---------------

inoperable during movement LCO 3.0.3 is not applicable.

of recently irradiated fuel ----------------------------------------

assemblies in the secondary containment or during F.1 Suspend movement of Immediately OPDRVs. recently irradiated fuel assemblies in the OR secondary containment.

One or more CREVAS AND subsystems inoperable due to an inoperable CRE F.2 Initiate action to suspend Immediately boundary during movement OPDRVs.

of recently irradiated fuel in the secondary containment or during OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Operate each CREVAS subsystem for 15 minutes. In accordance with the Surveillance Frequency Control Program SR 3.7.3.2 Perform required CREVAS filter testing in accordance In accordance with the Ventilation Filter Testing Program (VFTP). with the VFTP (continued)

JAFNPP 3.7.3-3 Amendment 301XXX

Control Room AC System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Air Conditioning (AC) System LCO 3.7.4 Two control room AC subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One control room AC A.1 Restore control room AC 30 days subsystem inoperable. subsystem to OPERABLE status.

B. Two control room AC B.1 Verify control room area Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months subsystems inoperable. temperature < 90 °F.

AND B.2 Restore one control 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months room AC subsystem to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A or B AND not met in MODE 1, 2, or

3. C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

JAFNPP 3.7.4-1 Amendment 290XXX

Control Room AC System 3.7.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and - - - - - - - - - NOTE - - - - - - - - - -

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A not ------------------------

met during movement of recently irradiated fuel D.1 Place OPERABLE Immediately assemblies in the control room AC secondary containment or subsystem in during OPDRVs. operation.

OR D.2.1 Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.

AND D.2.2 Initiate action to suspend Immediately OPDRVs.

E. Required Action and - - - - - - - - - NOTE - - - - - - - - - -

associated Completion LCO 3.0.3 is not applicable.

Time of Condition B not ------------------------

met during movement of recently irradiated fuel E.1 Suspend movement of Immediately assemblies in the recently irradiated fuel secondary containment or assemblies in the during OPDRVs. secondary containment.

AND E.2 Initiate action to suspend Immediately OPDRVs.

JAFNPP 3.7.4-2 Amendment 290XXX

AC Sources - Shutdown 3.8.2 ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or both required -------------------NOTE------------------

offsite circuits Enter applicable Condition inoperable. and Required Actions of LCO 3.8.8, when any required division is de-energized as a result of Condition A.

A.1 Declare affected Immediately required feature(s),

with no offsite power available, inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND A.2.2 Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.

AND A.2.3 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel (OPDRVs).

AND (continued)

JAFNPP 3.8.2-2 Amendment 276XXX

AC Sources - Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.34 Initiate action to Immediately restore required offsite power circuit(s) to OPERABLE status.

B. One required EDG B.1 Suspend CORE Immediately subsystem inoperable. ALTERATIONS.

AND B.2 Suspend movement of Immediately recently irradiated fuel assemblies in secondary containment.

AND B.3 Initiate action to Immediately suspend OPDRVs.

AND B.34 Initiate action to Immediately restore required EDG subsystem to OPERABLE status.

JAFNPP 3.8.2-3 Amendment 276XXX

AC Sources Shutdown 3.8.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.2.1 ---------------------------NOTES---------------------

1. The following SRs are not required to be performed: SR 3.8.1.3, SR 3.8.1.8, SR 3.8.1.9, SR 3.8.1.11, SR 3.8.1.12, and SR 3.8.1.13.

2. SR 3.8.1.10 and SR 3.8.1.12 are not required to be met when associated ECCS subsystem(s) are not required to be OPERABLE per LCO 3.5.2, "Reactor Pressure Vessel ECCS -

Shutdown(RPV) Water Inventory Control."

For AC sources required to be OPERABLE the In accordance SRs of Specification 3.8.1, except with applicable SR 3.8.1.7, are applicable. SRs JAFNPP 3.8.2-4 Amendment 274XXX

DC Sources - Shutdown 3.8.5 ACTIONS CONDITIONS REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel.

AND A.2.34 Initiate action to Immediately restore required DC electrical power subsystem to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.5.1 -----------------------NOTE----------------------

The following SRs are not required to be performed: SR 3.8.4.2, SR 3.8.4.3, and SR 3.8.4.4.

For DC electrical power subsystem required In accordance to be OPERABLE the following SRs are with applicable applicable: SRs SR 3.8.4.1, SR 3.8.4.2, SR 3.8.4.3, and SR 3.8.4.4.

JAFNPP 3.8.5-2 Amendment 274XXX

Distribution Systems - Shutdown 3.8.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A. (continued)

A.2.3 Initiate action to suspend Immediately operations with a potential for draining the reactor vessel.

AND A.2.34 Initiate actions to restore Immediately required AC and 125 VDC electrical power distribution subsystems to OPERABLE status.

AND A.2.45 Declare associated Immediately required shutdown cooling subsystem(s) inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and voltage to In accordance with required AC and 125 VDC electrical power distribution the Surveillance subsystems. Frequency Control Program JAFNPP 3.8.8-2 Amendment 301XXX

ATTACHMENT 3 Markup of Technical Specification Bases Pages James A. FitzPatrick Nuclear Power Station Renewed Facility Operating License Nos. DPR-59 Docket Nos. 50-333 Revised Technical Specification Bases Pages Unit 1 TS Bases Pages 8 3.3.5.3-9 8 3.6.4.1-2 ii 8 3.3.5.3-10 8 3.6.4.1-3 8 3.3.4.1-4 8 3.3.5.3-11 8 3.6.4.2-2 8 3.3.5.1-8 8 3.3.5.3-12 8 3.6.4.2-5 8 3.3.5.1-10 8 3.3.6.1-7 8 3.6.4.3-3 8 3.3.5.1-12 8 3.3.6.1-22 8 3.6.4.3-4 8 3.3.5.1-14 8 3.3.6.1-23 8 3.6.4.3-5 8 3.3.5.1-26 8 3.3.6.1-29 8 3.7.3-3 8 3.3.5.1-28 8 3.3.6.2-4 8 3.7.3-5 8 3.3.5.1-31 8 3.3.7.1-3 8 3.7.3-6 8 3.3.5.2-1* 8 3.5.1-1 8 3.7.4-2 8 3.3.5.2-2* 8 3.5.1-6 8 3.7.4-3 8 3.3.5.2-3* 8 3.5.2-1* 8 3.7.4-4 8 3.3.5.2-4* 8 3.5.2-2* 8 3.7.4-5 8 3.3.5.2-5* 8 3.5.2-3* 8 3.8.2-1 8 3.3.5.2-6* 8 3.5.2-4* 8 3.8.2-3 8 3.3.5.2-7* 8 3.5.2-5* 8 3.8.2-4 8 3.3.5.2-8* 83.5.2-6* 8 3.8.2-5 83.3.5.2-9* 8 3.5.2-7* 8 3.8.2-6 8 3.3.5.3-1 8 3.5.2-8* 8 3.8.5-1 8 3.3.5.3-2 8 3.5.2-9* 8 3.8.5-2 8 3.3.5.3-3 8 3.5.2-10* 8 3.8.5-3 8 3.3.5.3-4 8 3.5.3-1 8 3.8.5-4 8 3.3.5.3-5 8 3.5.3-2 8 3.8.8-1 8 3.3.5.3-6 8 3.6.1.3-4 8 3.8.8-2 8 3.3.5.3-7 8 3.6.1.3-10 8 3.8.8-3 8 3.3.5.3-8 8 3.6.2.2-2 8 3.10.1-2

New TS Page

TABLE OF CONTENTS B 2.0 SAFETY LIMITS (SLs)

B 2.1 Reactor Core SLs............................................................................................. B 2.1.1-1 B 2.2 Reactor Coolant System (RCS) Pressure SL ..................................................B 2.1.2-1 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY ........................B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY .......................................B 3.0-12 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM) ......................................................................... B 3.1.1-1 B 3.1.2 Reactivity Anomalies ........................................................................................ B 3.1.2-1 B 3.1.3 Control Rod OPERABILITY ............................................................................. B 3.1.3-1 B 3.1.4 Control Rod Scram Times................................................................................ B 3.1.4-1 B 3.1.5 Control Rod Scram Accumulators.................................................................... B 3.1.5-1 B 3.1.6 Rod Pattern Control ......................................................................................... B 3.1.6-1 B 3.1.7 Standby Liquid Control (SLC) System .............................................................. B 3.1.7-1 B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves .................................B 3.1.8-1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ..........B 3.2.1-1 B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)...............................................B 3.2.2-1 B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR).................................................B 3.2.3-1 B 3.3 INSTRUMENTATION B 3.3.1.1 Reactor Protection System (RPS) Instrumentation ..........................................B 3.3.1.1-1 B 3.3.1.2 Source Range Monitor (SRM) Instrumentation ................................................B 3.3.1.2-1 B 3.3.2.1 Control Rod Block Instrumentation .................................................................. B 3.3.2.1-1 B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation ................................................................................ B 3.3.2.2-1 B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation ............................................B 3.3.3.1-1 B 3.3.3.2 Remote Shutdown System ............................................................................... B 3.3.3.2-1 B 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation ...............................................B 3.3.4.1-1 B 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation .............................B 3.3.5.1-1 B 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation .....B 3.3.5.2-1 B 3.3.5.23 Reactor Core Isolation Cooling (RCIC) System Instrumentation ....................................................................................... B 3.3.5.23-1 B 3.3.6.1 Primary Containment Isolation Instrumentation ...............................................B 3.3.6.1-1 B 3.3.6.2 Secondary Containment Isolation Instrumentation ..........................................B 3.3.6.2-1 B 3.3.7.1 Control Room Emergency Ventilation Air Supply (CREVAS)

System Instrumentation .......................................................................... B 3.3.7.1-1 B 3.3.7.2 Condenser Air Removal Pump Isolation Instrumentation ................................B 3.3.7.2-1 B 3.3.7.3 Emergency Service Water (ESW) System Instrumentation ............................B 3.3.7.3-1 B 3.3.8.1 Loss of Power (LOP) Instrumentation .....................................................B 3.3.8.1-1 B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring ....................................................................................... B 3.3.8.1-2 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 Recirculation Loops Operating ................................................................ B 3.4.1-1 B 3.4.2 Jet Pumps ............................................................................................... B 3.4.2-1 (continued)

JAFNPP i Revision 18XX

TABLE OF CONTENTS B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

B 3.4.3 Safety/Relief Valves (S/RVs) ................................................................... B 3.4.3-1 B 3.4.4 RCS Operational LEAKAGE ................................................................... B 3.4.4-1 B 3.4.5 RCS Leakage Detection Instrumentation ................................................B 3.4.5-1 B 3.4.6 RCS Specific Activity ............................................................................... B 3.4.6-1 B 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown ................................................................. B 3.4.7-1 B 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown ............................................................... B 3.4.8-1 B 3.4.9 RCS Pressure and Temperature (P/T) Limits .........................................B 3.4.9-1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS - Operating................................................................................... B 3.5.1-1 B 3.5.2 RPV Water Inventory ControlECCS - Shutdown ....................................B 3.5.2-1 B 3.5.3 RCIC System ........................................................................................... B 3.5.3-1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment............................................................................... B 3.6.1.1-1 B 3.6.1.2 Primary Containment Air Locks ............................................................... B 3.6.1.2-1 B 3.6.1.3 Primary Containment Isolation Valves (PCIVs) .......................................B 3.6.1.3-1 B 3.6.1.4 Drywell Pressure ..................................................................................... B 3.6.1.4-1 B 3.6.1.5 Drywell Air Temperature.......................................................................... B 3.6.1.5-1 B 3.6.1.6 Reactor Building-to-Suppression Chamber Vacuum Breakers ........................................................................... B 3.6.1.6-1 B 3.6.1.7 Suppression Chamber-to-Drywell Vacuum Breakers..............................B 3.6.1.7-1 B 3.6.1.8 Main Steam Leakage Collection (MSLC) System ...................................B 3.6.1.8-1 B 3.6.1.9 Residual Heat Removal (RHR) Containment Spray System ..................B 3.6.1.9-1 B 3.6.2.1 Suppression Pool Average Temperature ................................................B 3.6.2.1-1 B 3.6.2.2 Suppression Pool Water Level ................................................................ B 3.6.2.2-1 B 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling .....................B 3.6.2.3-1 B 3.6.2.4 Drywell-to-Suppression Chamber Differential Pressure..........................B 3.6.2.4-1 B 3.6.3.1 Primary Containment Oxygen Concentration ..........................................B 3.6.3.1-1 B 3.6.3.2 Containment Atmosphere Dilution (CAD) System ..................................B 3.6.3.2-1 B 3.6.4.1 Secondary Containment .......................................................................... B 3.6.4.1-1 B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) ..................................B 3.6.4.2-1 B 3.6.4.3 Standby Gas Treatment (SGT) System ..................................................B 3.6.4.3-1 B 3.7 PLANT SYSTEMS B 3.7.1 Residual Heat Removal Service Water (RHRSW) System ....................B 3.7.1-1 B 3.7.2 Emergency Service Water (ESW) System and Ultimate Heat Sink (UHS) ............................................................................. B 3.7.2-1 B 3.7.3 Control Room Emergency Ventilation Air Supply (CREVAS) System .............................................................. B 3.7.3-1 B 3.7.4 Control Room Air Conditioning (AC) System ..........................................B 3.7.4-1 B 3.7.5 Main Condenser Steam Jet Air Ejector (SJAE) System ..........................B 3.7.5-1 (continued)

JAFNPP ii Revision 18XX

ATWS RPT Instrumentation B 3.3.4.1 BASES APPLICABLE a. Reactor Vessel Water Level - Low Low (Level 2)

SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY Four channels of Reactor Vessel Water Level - Low Low (Level 2), with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Vessel Water Level - Low Low (Level 2) Allowable Value is chosen so that the system will not be initiated after a Level 3 scram with feedwater still available, and also provides an opportunity for the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems to recover water level if feedwater is not available. The Allowable Value is referenced from a level of water 352.56 inches above the lowest point in the inside bottom of the RPV and also corresponds to the top of a 144 inch fuel column (Ref. 3).

The HPCI, RCIC and ATWS-RPT initiation functions (as described in Table 3.3.5.1-1, Function 3.a; Table 3.3.5.23-1. Function 1; and LCO 3.3.4.1.a including SR 3.3.4.1.4. respectively) describe the reactor vessel water level initiation function as "Low Low (Level 2)." The Allowable Values associated with the HPCI and RCIC initiation function is different from the Allowable Value associated with the ATWS-RPT initiation function as the ATWS function has a separate analog trip unit. Nevertheless, consistent with the nomenclature typically used in design documents, the "Low Low (Level 2)" designation is retained in describing each of these three initiation functions.

b. Reactor Pressure - High Excessively high RPV pressure may rupture the RCPB.

An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Pressure-High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT (continued)

JAFNPP B 3.3.4.1-4 Revision 0XX

ECCS Instrumentation B 3.3.5.1 BASES BACKGROUND Emergency Diesel Generators (continued) associated ECCS initiation logic. Upon receipt of an ECCS initiation signal, each EDG is automatically started, is ready to load in approximately 10 seconds, and will run in standby conditions (rated voltage and speed, with the EDG output breaker open). The EDGs will only energize their respective emergency buses if a loss of preferred power occurs. (Refer to Bases for LCO 3.3.8.1.)

APPLICABLE The actions of the ECCS are explicitly assumed in the safety SAFETY ANALYSES, analyses of References 1, 2, 3, and 4. The ECCS is LCO, and initiated to preserve the integrity of the fuel cladding by APPLICABILITY limiting the post LOCA peak cladding temperature to less than the 10 CFR 50.46 limits.

ECCS instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 5). Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.

The OPERABILITY of the ECCS instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.

Table 3.3.5.1-1 is modified by two a footnotes which.

Footnote (a) is added to clarify that the associated functions are required to be OPERABLE in MODES 4 and 5 only when their supported ECCS are required to be OPERABLE per LCO 3.5.2, "ECCSShutdown." Footnote (b) is added to show that certain ECCS instrumentation Functions also perform EDG initiation.

Allowable Values are specified for each ECCS Function specified in the table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS.

Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water (continued)

JAFNPP B 3.3.5.1-8 Revision 0XX

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 1.a, 2.a. Reactor Vessel Water Level Low Low Low (Level 1)

SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY cooling function of the ECCS, along with the scram action of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

Reactor Vessel Water Level Low Low Low (Level 1) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

The Reactor Vessel Water Level Low Low Low (Level 1)

Allowable Value is chosen to allow time for the low pressure core flooding systems to activate and provide adequate cooling. The Allowable Value is referenced from a level of water 352.56 inches above the lowest point in the inside bottom of the RPV and also corresponds to the top of a 144 inch fuel column (Ref. 6).

Thus, four channels of the CS and LPCI Reactor Vessel Water Level Low Low Low (Level 1) Function are only required to be OPERABLE when the ECCS are required to be OPERABLE to ensure that no single instrument failure can preclude ECCS initiation. Per Footnote (a) to Table 3.3.5.1-1, this ECCS Function is only required to be OPERABLE in MODES 4 and 5 whenever the associated ECCS is required to be OPERABLE per LCO 3.5.2. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems; LCO 3.8.1 and LCO 3.8.2, "AC SourcesShutdown," for Applicability Bases for the EDGs.

1.b, 2.b. Drywell Pressure High High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB). The low pressure ECCS and associated EDGs are initiated upon receipt of the Drywell Pressure High Function in order to minimize the possibility of fuel damage. The EDGs are initiated from Function 1.b and 2.b. The Drywell Pressure High Function, along with the Reactor Water Level Low Low Low (Level 1)

Function, is directly assumed in the analysis of the recirculation line break (Refs. 1, 2, and 4). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

(continued)

JAFNPP B 3.3.5.1-10 Revision 0XX

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 1.c, 2.c. Reactor Pressure Low (Injection Permissive)

SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY Four channels of Reactor Pressure Low Function are only required to be OPERABLE when the ECCS is required to be OPERABLE to ensure that no single instrument failure can preclude ECCS initiation. Per Footnote (a) to Table 3.3.5.1-1, this ECCS Function is only required to be OPERABLE in MODES 4 and 5 whenever the associated ECCS is required to be OPERABLE per LCO 3.5.2. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.

1.d, 2.f. Core Spray and Low Pressure Coolant Injection Pump Start-Time Delay Relay The purpose of these time delay relays is to stagger the start of the CS and LPCI pumps to enable sequential loading of the appropriate AC source. The CS and LPCI Pump Start-Time Delay Relays are assumed to be OPERABLE in the accident analyses requiring ECCS initiation. That is, the analyses assumes that the pumps will initiate when required and no excess loading of the power sources will occur.

There are two CS and four LPCI Pump Start-Time Delay Relays, one in each of the CS and LPCI pump start circuits. While each time delay relay is dedicated to a single pump start circuit, a single failure of a CS or LPCI Pump Start-Time Delay Relay could result in the failure of a CS pump and both the LPCI pumps powered from the same emergency bus to perform their intended function within the assumed ECCS response time (e.g., as in the case where one inoperable time delay relay results in more than one pump starting at nearly the same time). In the worst case this would still leave the other three low pressure ECCS pumps OPERABLE; thus, the single failure of one instrument does not preclude ECCS initiation. The Allowable Values for the CS and LPCI Pump Start-Time Delay Relays are chosen to be short enough so that ECCS operation is within the time period assumed in the accident analyses.

Each CS and LPCI Pump Start-Time Delay Relay Function is required to be OPERABLE only when the associated CS and LPCI subsystem is required to be OPERABLE. Per Footnote (a) to Table 3.3.5.1-1, this ECCS Function is only required to be OPERABLE in MODES 4 and 5 whenever the associated ECCS is required to be OPERABLE per LCO 3.5.2. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the CS and LPCI subsystems.

(continued)

JAFNPP B 3.3.5.1-12 Revision 0XX

ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 1.e, 2.g, 1.f. Core Spray and Low Pressure Coolant SAFETY ANALYSES, Injection Pump Discharge Flow Low (Bypass), Core Spray LCO, and Pump Discharge Pressure High (Bypass) (continued)

APPLICABILITY condition that results in a discharge pressure permissive when the CS pump is aligned for injection and the pump is not running.

Each channel of Pump Discharge FlowLow Function (two CS channels and four LPCI channels) and each channel of Core Spray Pump Discharge PressureHigh (Bypass) are only required to be OPERABLE when the associated ECCS is required to be OPERABLE to ensure that no single instrument failure can preclude the ECCS function. Per Footnote (a) to Table 3.3.5.1-1, this ECCS Function is only required to be OPERABLE in MODES 4 and 5 whenever the associated ECCS is required to be OPERABLE per LCO 3.5.2. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.

2.d. Reactor Pressure Low (Recirculation Discharge Valve Permissive)

Low reactor pressure signals are used as permissives for recirculation discharge valve closure. This ensures that the LPCI subsystems inject into the proper RPV location assumed in the safety analysis. The Reactor Pressure Low is one of the Functions assumed to be OPERABLE and capable of closing the valve during the transients analyzed in Reference 3. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The Reactor Pressure Low Function is directly assumed in the analysis of the recirculation line break (Refs. 1, 2 and 4).

The Reactor Pressure Low signals are initiated from four pressure transmitters that sense the reactor dome pressure.

The Allowable Value is chosen to ensure that the valves close prior to commencement of LPCI injection flow into the core, as assumed in the safety analysis.

Four channels of the Reactor PressureLow Function are only required to be OPERABLE in MODES 1, 2, and 3 with the associated recirculation pump discharge valve open. With the valve(s) closed, the function of the instrumentation has been performed; thus, the Function is not required. In (continued)

JAFNPP B 3.3.5.1-14 Revision 0XX

ECCS Instrumentation B 3.3.5.1 BASES ACTIONS B.1, B.2, and B.3 (continued) more Function 2.a channels are inoperable and untripped such that both trip systems lose initiation capability, (c) two or more Function 1.b channels are inoperable and untripped such that both trip systems lose initiation capability, or (d) two or more Function 2.b channels are inoperable and untripped such that both trip systems lose initiation capability. For low pressure ECCS, since each inoperable channel would have Required Action B.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected portion of the associated system of low pressure ECCS and EDGs to be declared inoperable. However, since channels in both associated low pressure ECCS subsystems (e.g., both CS subsystems) are inoperable and untripped, and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in the associated low pressure ECCS and EDGs being concurrently declared inoperable.

For Required Action B.2, redundant automatic HPCI initiation capability is lost if two or more Function 3.a or two or more Function 3.b channels are inoperable and untripped such that trip capability is lost. In this situation (loss of redundant automatic initiation capability), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Action B.3 is not appropriate and the HPCI System must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . As noted (Note 1 to Required Action B.1), Required Action B.1 is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the low pressure ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of initiation capability for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (as allowed by Required Action B.3) is allowed during MODES 4 and 5.

There is no similar Note provided for Required Action B.2 since HPCI instrumentation is not required in MODES 4 and 5; thus, a Note is not necessary.

Notes are also provided (the Note 2 to Required Action B.1 and the Note to Required Action B.2) to delineate which Required Action is applicable for each Function that requires entry into Condition B if an associated channel is inoperable.

This ensures that the proper loss of initiation capability check is performed. Required Action B.1 (the Required Action for certain inoperable channels in the low pressure ECCS subsystems) is not applicable to Functions 2.e and 2.h, since these Functions provide backup to administrative controls ensuring that operators do not divert LPCI flow from injecting into the core when needed, and do not spray the containment unless needed. Thus, a total loss of (continued)

JAFNPP B 3.3.5.1-26 Revision 0XX

ECCS Instrumentation B 3.3.5.1 BASES ACTIONS C.1 and C.2 (continued)

ECCS). Redundant automatic initiation capability is lost if either (a) two or more Function 1.c channels are inoperable such that both trip systems lose initiation capability, (b) two Function 1.d channels are inoperable, (c) two or more Function 2.c channels are inoperable such that both trip systems lose initiation capability, (d) two or more Function 2.d channels are inoperable such that both trip systems lose initiation capability, or (e) three Function 2.f channels are inoperable. In this situation (loss of redundant automatic initiation capability), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Action C.2 is not appropriate and the feature(s) associated with the inoperable channels must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Since each inoperable channel would have Required Action C.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected portion of the associated system to be declared inoperable. However, since channels for both low pressure ECCS subsystems are inoperable (e.g.,

both CS subsystems), and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in both subsystems being concurrently declared inoperable. For Functions 1.c, 1.d, 2.c, 2.d, and 2.f, the affected portions are the associated low pressure ECCS pumps. As noted (Note 1), Required Action C.1 is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of automatic initiation capability for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (as allowed by Required Action C.2) is allowed during MODES 4 and 5.

The Note 2 states that Required Action C.1 is only applicable for Functions 1.c, 1.d, 2.c, 2.d, and 2.f.

Required Action C.1 is not applicable to Function 3.c (which also requires entry into this Condition if a channel in this Function is inoperable), since the loss of one channel results in a loss of the Function (two out of two logic).

This loss was considered during the development of Reference 7 and considered acceptable for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed by Required Action C.2.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."

For Required Action C.1, the Completion Time only begins upon discovery that the same feature in both subsystems (continued)

JAFNPP B 3.3.5.1-28 Revision 0XX

ECCS Instrumentation B 3.3.5.1 BASES ACTIONS E.1 and E.2 (continued) concurrently for the channels of the low pressure ECCS pumps, this results in the affected low pressure ECCS pumps being concurrently declared inoperable.

In this situation (loss of redundant automatic initiation capability), the 7 day allowance of Required Action E.2 is not appropriate and the subsystem associated with each inoperable channel must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . As noted (Note 1 to Required Action E.1), Required Action E.1 is only applicable in MODES 1. 2. and 3. In MODES 4 and 5, the specific initiation time of the ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of initiation capability for 7 days (as allowed by Required Action E.2) is allowed during MODES 4 and 5. A Note is also provided (the Note 2 to Required Action E.1) to delineate that Required Action E.1 is only applicable to low pressure ECCS Functions. Required Action E.1 is not applicable to HPCI Functions 3.f and 3.g since the loss of one channel results in a loss of the Function (one-out-of-one logic). This loss was considered during the development of Reference 7 and considered acceptable for the 7 days allowed by Required Action E.2.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."

For Required Action E.1, the Completion Time only begins upon discovery that a redundant feature in the same system (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.

If the instrumentation that controls the pump minimum flow valve is inoperable, such that the valve will not automatically open, extended pump operation with no injection path available could lead to pump overheating and failure. If there were a failure of the instrumentation.

such that the valve would not automatically close, a portion of the pump flow could be diverted from the reactor vessel injection path, causing insufficient core cooling. These consequences can be averted by the operator's manual control of the valve, which would be adequate to maintain ECCS pump (continued)

JAFNPP B 3.3.5.1-31 Revision 0XX

RPV Water Inventory Control Instrumentation B 3.3.5.2 B 3.3 INSTRUMENTATION B 3.3.5.2 REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL INSTRUMENTATION BASES BACKGROUND The RPV contains penetrations below the top of the active fuel (TAF) that have the potential to drain the reactor coolant inventory to below the TAF. If the water level should drop below the TAF, the ability to remove decay heat is reduced, which could lead to elevated cladding temperatures and clad perforation. Safety Limit 2.1.1.3 requires the RPV water level to be above the top of the active irradiated fuel at all times to prevent such elevated cladding temperatures.

Technical Specifications are required by 10 CFR 50.36 to include limiting safety system settings (LSSS) for variables that have significant safety functions. LSSS are defined by the regulation as "Where a LSSS is specified for a variable on which a safety limit has been placed, the setting must be chosen so that automatic protective actions will correct the abnormal situation before a Safety Limit (SL) is exceeded."

The Analytical Limit is the limit of the process variable at which a safety action is initiated to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the Analytical Limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protection channels must be chosen to be more conservative than the Analytical Limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur.

The actual settings for the automatic isolation channels are the same as those established for the same functions in MODES 1, 2, and 3 in LCO 3.3.5.1, "Emergency Core Cooling System (ECCS) Instrumentation," or LCO 3.3.6.1, "Primary Containment Isolation instrumentation".

With the unit in MODE 4 or 5, RPV water inventory control is not required to mitigate any events or accidents evaluated in the safety analyses. RPV water inventory control is required in MODES 4 and 5 to protect Safety Limit 2.1.1.3 and the fuel cladding barrier to prevent the release of radioactive material should a draining event occur. Under the definition of DRAIN TIME, some penetration flow paths may be excluded from the DRAIN TIME calculation if they will be isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation.

(continued)

JAFNPP B 3.3.5.2-1 Revision XX

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES BACKGROUND The purpose of the RPV Water Inventory Control (continued) Instrumentation is to support the requirements of LCO 3.5.2, Reactor Pressure Vessel (RPV) Water Inventory Control, and the definition of DRAIN TIME. There are functions that are required for manual initiation or operation of the ECCS injection/spray subsystem required to be OPERABLE by LCO 3.5.2 and other functions that support automatic isolation of Residual Heat Removal subsystem and Reactor Water Cleanup system penetration flow path(s) on low RPV water level.

The RPV Water Inventory Control Instrumentation supports operation of core spray (CS) and low pressure coolant injection (LPCI). The equipment involved with each of these systems is described in the Bases for LCO 3.5.2.

APPLICABLE With the unit in MODE 4 or 5, RPV water inventory control is SAFETY ANALYSES, not required to mitigate any events or accidents evaluated LCO, and in the safety analyses. RPV, LCO, water inventory control is APPLICABILITY required in MODES 4 and 5 to protect and Safety Limit 2.1.1.3 and the fuel cladding barrier to prevent the release of radioactive material should a draining event occur.

A double-ended guillotine break of the Reactor Coolant System (RCS) is not postulated in MODES 4 and 5 due to the reduced RCS pressure, reduced piping stresses, and ductile piping systems. Instead, an event is postulated in which a single operator error or initiating event allows draining of the RPV water inventory through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e.g., seismic event, loss of normal power, single human error). It is assumed, based on engineering judgment, that while in MODES 4 and 5, one low pressure ECCS injection/spray subsystem can be manually initiated to maintain adequate reactor vessel water level.

As discussed in References 1, 2, 3, 4, and 5, operating experience has shown RPV water inventory to be significant to public health and safety. Therefore, RPV Water Inventory Control satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

Permissive and interlock setpoints are generally considered as nominal values without regard to measurement accuracy.

(continued)

JAFNPP B 3.3.5.2-2 Revision XX

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES APPLICABLE Core Spray and Low Pressure Coolant Injection Systems SAFETY ANALYSES, LCO, and 1.a, 2.a. Reactor Pressure - Low (Injection Permissive)

APPLICABILITY (continued) Low reactor pressure signals are used as permissives for the low pressure ECCS injection/spray subsystem manual injection functions. This function ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems' maximum design pressure. While it is assured during MODES 4 and 5 that the reactor pressure will be below the ECCS maximum design pressure, the Reactor Pressure - Low signals are assumed to be OPERABLE and capable of permitting initiation of the ECCS. The Reactor Pressure - Low signals are initiated from four pressure transmitters that sense the reactor dome pressure. The Allowable Value is low enough to prevent overpressuring the equipment in the low pressure ECCS.

The four channels of Reactor Pressure - Low Function are required to be OPERABLE in MODES 4 and 5 when ECCS manual initiation is required to be OPERABLE by LCO 3.5.2.

1.b, 2.b, and 1.c. Core Spray and Low Pressure Coolant Injection Pump Discharge Flow Low (Bypass), Core Spray Pump Discharge Pressure High (Bypass) (continued)

The minimum flow instruments are provided to protect the associated low pressure ECCS pump from overheating when the pump is operating and the associated injection valve is not fully open. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump.

One differential pressure indicating switch per CS pump and one differential pressure indicating switch per LPCI subsystem are used to detect the associated subsystems' flow rates. In addition, one pressure switch per CS pump is used to detect the associated pumps discharge pressure. The logic is arranged such that each differential pressure indicating switch causes its associated minimum flow valve to open. For CS both the differential pressure indicating switch and the pressure switch must actuate to cause the valve to open. The logic will close the minimum flow valve once the closure setpoint of the associated differential pressure indicating switch is exceeded. The LPCI minimum flow valves are time delayed such that the valves will not open for 10 seconds after the switches detect low flow. The time delay is provided to limit reactor vessel inventory loss during the startup of the Residual Heat Removal (RHR) shutdown cooling mode.

(continued)

JAFNPP B 3.3.5.2-3 Revision XX

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES APPLICABLE The Pump Discharge Flow - Low Allowable Values are high SAFETY ANALYSES, enough to ensure that the pump flow rate is sufficient to LCO, and protect the pump, yet low enough to ensure that the closure APPLICABILITY of the minimum flow valve is initiated to allow full flow (continued) into the core. The Core Spray Pump Discharge Pressure-High (Bypass) Allowable Value is less than the pump discharge pressure when the pump is operating in a full flow mode and high enough to avoid any condition that results in a discharge pressure permissive when the CS pump is aligned for injection and the pump is not running.

One channel of the Pump Discharge Flow - Low Function and one channel of Core Spray Pump Discharge Pressure - High (Bypass) is required to be OPERABLE in MODES 4 and 5 when the associated Core Spray or LPCI pump is required to be OPERABLE by LCO 3.5.2 to ensure the pumps are capable of injecting into the Reactor Pressure Vessel when manually initiated.

RHR System Isolation 3.a - Reactor Vessel Water Level - Low, Level 3 The definition of Drain Time allows crediting the closing of penetration flow paths that are capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation. The Reactor Vessel Water Level - Low, Level 3 Function associated with RHR System isolation may be credited for automatic isolation of penetration flow paths associated with the RHR System.

Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. While four channels (two channels per trip system) of the Reactor Vessel Water Level - Low, Level 3 Function are available, only two channels (all in the same trip system) are required to be OPERABLE.

(continued)

JAFNPP B 3.3.5.2-4 Revision XX

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES APPLICABLE The Reactor Vessel Water Level - Low, Level 3 Allowable SAFETY ANALYSES, Value was chosen to be the same as the Primary Containment LCO, and Isolation Instrumentation Reactor Vessel Water Level - Low, APPLICABILITY Level 3 Allowable Value (LCO 3.3.6.1), since the capability (continued) to cool the fuel may be threatened.

The Reactor Vessel Water Level - Low, Level 3 Function is only required to be OPERABLE when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME. This Function isolates the Group 11 valves.

Reactor Water Cleanup (RWCU) System Isolation 4.a - Reactor Vessel Water level - Low, Level 3 The definition of Drain Time allows crediting the closing of penetration flow paths that are capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation. The Reactor Vessel Water Level - Low (Level 3) Function associated with RWCU System isolation may be credited for automatic isolation of penetration flow paths associated with the RWCU System.

Reactor Vessel Water Level - Low, (Level 3) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. While four channels (two channels per trip system) of the Reactor Vessel Water Level - Low (Level 3) Function are available, only two channels (all in the same trip system) are required to be OPERABLE.

The Reactor Vessel Water Level - Low (Level 3) Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level - Low (Level 3) Allowable Value (LCO 3.3.1.1),

since the capability to cool the fuel may be threatened. The allowable value is referenced from a level of water 352.56 inches above thr lowest point in the inside bottom of the RPV and also corresponds to the top of a 144 inch fuel column (Ref 6).

The Reactor Vessel Water Level - Low (Level 3) Function is only required to be OPERABLE when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME. This Function isolates the Group 5 valves.

(continued)

JAFNPP B 3.3.5.2-5 Revision XX

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES ACTIONS A Note has been provided to modify the ACTIONS related to RPV Water Inventory Control instrumentation channels.

Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition.

Section 1.3 also specifies that Required Actions continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RPV Water Inventory Control instrumentation channels provide appropriate compensatory measures for separate inoperable Condition entry for each inoperable RPV Water Inventory Control instrumentation channel.

A.1 Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.2-1. The applicable Condition referenced in the Table is Function dependent.

Each time a channel is discovered inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.

B.1 and B.2 RHR System Isolation, Reactor Vessel Water Level - Low Level 3, and Reactor Water Cleanup System, Reactor Vessel Water Level - Low Low, Level 2 functions are applicable when automatic isolation of the associated penetration flow path is credited in calculating Drain Time. If the instrumentation is inoperable, Required Action B.1 directs an immediate declaration that the associated penetration flow path(s) are incapable of automatic isolation. Required Action B.2 directs calculation of DRAIN TIME. The calculation cannot credit automatic isolation of the affected penetration flow paths.

(continued)

JAFNPP B 3.3.5.2-6 Revision XX

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES ACTIONS C.1 (continued)

Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS injection/spray subsystem manual injection functions. If the permissive is inoperable, manual initiation of ECCS is prohibited.

Therefore, the permissive must be placed in the trip condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . With the permissive in the trip condition, manual initiation may be performed. Prior to placing the permissive in the tripped condition, the operator can take manual control of the pump and the injection valve to inject water into the RPV.

The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is intended to allow the operator time to evaluate any discovered inoperabilities and to place the channel in trip.

D.1 If a Core Spray or Low Pressure Coolant Injection Pump Discharge Flow - Low bypass function is inoperable, there is a risk that the associated low pressure ECCS pump could overheat when the pump is operating and the associated injection valve is not fully open. In this condition, the operator can take manual control of the pump and the injection valve to ensure the pump does not overheat. If a manual initiation function is inoperable, the ECCS subsystem pumps can be started manually and the valves can be opened manually, but this is not the preferred condition.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time was chosen to allow time for the operator to evaluate and repair any discovered inoperabilities. The Completion Time is appropriate given the ability to manually start the ECCS pumps and open the injection valves and to manually ensure the pump does not overheat.

E.1 With the Required Action and associated Completion Time of Condition C or D not met, the associated low pressure ECCS injection/spray subsystem may be incapable of performing the intended function, and must be declared inoperable immediately.

(continued)

JAFNPP B 3.3.5.2-7 Revision XX

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES SURVEILLANCE As noted in the beginning of the SRs, the SRs for each RPV REQUIREMENTS Water Inventory Control instrument Function are found in the SRs column of Table 3.3.5.2-1.

SR 3.3.5.2.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.

Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK guarantees that undetected outright channel failure is limited; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL FUNCTIONAL TEST. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

(continued)

JAFNPP B 3.3.5.2-8 Revision XX

RPV Water Inventory Control Instrumentation B 3.3.5.2 BASES SURVEILLANCE SR 3.3.5.2.2 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests.

Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.5.2.3 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.5.2 overlaps this Surveillance to complete testing of the assumed safety function.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. Information Notice 84-81 "Inadvertent Reduction in Primary Coolant Inventory in Boiling Water Reactors During Shutdown and Startup," November 1984.

2. Information Notice 86-74, "Reduction of Reactor Coolant Inventory Because of Misalignment of RHR Valves," August 1986.

3. Generic Letter 92-04, "Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(F), " August 1992.

4. NRC Bulletin 93-03, "Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs," May 1993.

5. Information Notice 94-52, "Inadvertent Containment Spray and Reactor Vessel Draindown at Millstone 1," July 1994.

6. Drawing 11825-5.01-150. Rev. D. Reactor Assembly Nuclear Boiler. (GE Drawing 9190690BO).

JAFNPP B 3.3.5.2-9 Revision XX

RCIC System Instrumentation 3.3.5.23 B 3.3 INSTRUMENTATION Reactor Core Isolation Cooling (RCIC) System Instrumentation BASES BACKGROUND The purpose of the RCIC System instrumentation is to initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink (the main condenser) and normal coolant makeup flow from the Reactor Feedwater System is insufficient or unavailable, such that RCIC System initiation occurs and maintains sufficient reactor water level such that an initiation of the low pressure Emergency Core Cooling Systems (ECCS) pumps does not occur. A more complete discussion of RCIC System operation is provided in the Bases of LCD 3.5.3, "RCIC System."

The RCIC System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level - Low Low (Level 2). The variable is monitored by four transmitters that are connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic arrangement. Once initiated, the RCIC logic seals in and can be reset by the operator only when the reactor vessel water level signals have cleared.

The normally closed RCIC test line isolation valve is closed on a RCIC initiation signal to allow full system flow.

The RCIC System also monitors the water level in each condensate storage tank (CST) since this is the initial source of water for RCIC operation. Reactor grade water in the CSTs is the normal source. The CST suction source consists of two CSTs connected in parallel to the RCIC pump suction. Upon receipt of a RCIC initiation signal, the CSTs suction valve is automatically signaled to open (it is normally in the open position) unless the pump suction from the suppression pool valves are open. If the water level in both CSTs fall below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve automatically closes. Two level switches are used to detect low water level in each CST. A level switch associated with each CST must actuate to cause the suppression pool suction valves to open and the CSTs suction valve to close. The channels are arranged in a one out-of-two taken twice logic. To prevent losing suction to the pump when automatically transferring suction from the (continued)

JAFNPP B 3.3.5.23-1 Revision 0XX

RCIC System Instrumentation B 3.3.5.23 BASES BACKGROUND CSTs to the suppression pool on low CST level. the suction (continued) valves are interlocked so that the suppression pool suction path must be open before the CST suction path automatically closes.

The RCIC System provides makeup water to the reactor until the reactor vessel water level reaches the high water level (Level 8) trip (two-out-of-two logic), at which time the RCIC steam inlet valve closes. The RCIC System restarts if vessel level again drops to the low level initiation point (Level 2).

APPLICABLE The function of the RCIC System is to respond to transient SAFETY ANALYSES, events by providing makeup coolant to the reactor. The LCO, and RCIC System is not an Engineered Safeguard System and APPLICABILITY no credit is taken in the safety analyses for RCIC System operation. The RCIC System instrumentation satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii) (Ref. 1). Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.

The OPERABILITY of the RCIC System instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.23-1. Each Function must have a required number of OPERABLE channels with their setpoints within the specified Allowable Values, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.

Allowable Values are specified for each RCIC System instrumentation Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations.

The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis or (continued)

JAFNPP B 3.3.5.23-2 Revision 0XX

RCIC System Instrumentation B 3.3.5.23 BASES APPLICABLE other appropriate documents. The trip setpoints are derived SAFETY ANALYSES, from the analytical limits and account for all worst case LCO, and instrumentation uncertainties as appropriate (e.g., drift, APPLICABILITY process effects, calibration uncertainties, and severe (continued) environmental errors (for channels that must function in harsh environments as defined by 10 CFR 50.49)). The trip setpoints derived in this manner provide adequate protection because all expected uncertainties are accounted for. The Allowable Values are then derived from the trip setpoints by accounting for normal effects that would be seen during periodic surveillance or calibration. These effects are instrumentation uncertainties observed during normal operation (e.g., drift and calibration uncertainties).

The individual Functions are required to be OPERABLE in MODE 1, and in MODES 2 and 3 with reactor steam dome pressure > 150 psig since this is when RCIC is required to be OPERABLE. (Refer to LCO 3.5.3 for Applicability Bases for the RCIC System.)

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.

1. Reactor Vessel Water Level - Low Low (Level 2)

Low reactor pressure vessel (RPV) water level indicates that normal feedwater flow is insufficient to maintain reactor vessel water level and that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the RCIC System is initiated at Level 2 to assist in maintaining water level above the top of the active fuel.

Reactor Vessel Water Level - Low Low (Level 2) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

The Reactor Vessel Water Level - Low Low (Level 2) Allowable Value is set high enough such that for complete loss of feedwater flow, the RCIC System flow with high pressure coolant injection assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Level 1. The (continued)

JAFNPP B 3.3.5.23-3 Revision 0XX

RCIC System Instrumentation B 3.3.5.23 BASES APPLICABLE 1. Reactor Vessel Water Level - Low Low (Level 2)

SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY Allowable Value is referenced from a level of water 352.56 inches above the lowest point in the inside bottom of the RPV and also corresponds to the top of a 144 inch fuel column (Ref. 2).

The HPCI. RCIC and ATWS-RPT initiation functions (as described in Table 3.3.5.1-1. Function 3.a; Table 3.3.5.23-1.

Function 1; and LCO 3.3.4.1.a including SR 3.3.4.1.4, respectively) describe the reactor vessel water level initiation function as "Low Low (Level 2)." The Allowable Values associated with the HPCI and RCIC initiation function is different from the Allowable Value associated with the ATWS*RPT initiation function as the ATWS function has a separate analog trip unit. Nevertheless, consistent with the nomenclature typically used in design documents, the "Low Low (Level 2)" is retained in describing each of these three initiation functions.

Four channels of Reactor Vessel Water Level - Low Low (Level 2) Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation.

Refer to LCO 3.5.3 for RCTC Applicability Bases.

2. Reactor Vessel Water Level-High (Level 8)

High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel. Therefore, the Level 8 signal is used to close the RCIC steam inlet valve to prevent overflow into the main steam lines (MSLs).

Reactor Vessel Water Level - High (Level 8) signals for RCIC are initiated from two level transmitters from the narrow range water level measurement instrumentation, which sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Both Level 8 signals are required in order to close the RCIC steam inlet valve.

The Reactor Vessel Water Level - High (Level 8) Allowable Value is high enough to preclude isolating the steam inlet valve during normal operation, yet low enough to prevent water overflowing into the MSLs. The Allowable Value is (continued)

JAFNPP B 3.3.5.23-4 Revision XXX0

RCIC System Instrumentation B 3.3.5.23 BASES APPLICABLE 2. Reactor Vessel Water Level - High (Level 8) (continued)

SAFETY ANALYSES, LCO, and referenced from a level of water 352.56 inches above the APPLICABILITY lowest point in the inside bottom of the RPV and also corresponds to the top of a 144 inch fuel column (Ref. 2).

Two channels of Reactor Vessel Water Level - High (Level 8)

Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases.

3. Condensate Storage Tank (CST) Level - Low Low level in the CSTs indicates the unavailability of an adequate supply of makeup water from this normal source.

Normally, the suction valve between the RCIC pump and the CSTs is open and, upon receiving a RCIC initiation signal, water for RCIC injection would be taken from the CSTs.

However, if the water level in both CSTs falls below a preselected level, first the suppression pool suction valves automatically open, and then the CSTs suction valve automatically closes. This ensures that an adequate supply of makeup water is available to the RCIC pump. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CSTs suction valve automatically closes.

Two level switches are used to detect low water level in each CST. The Condensate Storage Tank Level-Low Function Allowable Value is set high enough (15,600 gallons of water is available in each CST) to ensure adequate pump suction head while water is being taken from the CST.

Four channels of Condensate Storage Tank Level-Low Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC automatic suction source alignment to suppression pool source. Refer to LCO 3.5.3 for RCIC Applicability Bases.

4. Manual Initiation The Manual Initiation push button switch introduces a signal into the RCIC System initiation logic that is redundant to the automatic protective instrumentation and provides manual initiation capability. There is one push button for the RCIC System.

(continued)

JAFNPP B 3.3.5.23-5 Revision 0XX

RCIC System Instrumentation B 3.3.5.23 BASES APPLICABLE 4. Manual Initiation (continued)

SAFETY ANALYSES, LCO, and The Manual Initiation Function is not assumed in any APPLICABILITY accident or transient analyses in the UFSAR. However, the Function is retained for overall redundancy and diversity of the RCIC function as required by the NRC in the plant licensing basis.

There is no Allowable Value for this Function since the channel is mechanically actuated based solely on the position of the push button. One channel of Manual Initiation is required to be OPERABLE when RCIC is required to be OPERABLE.

ACTIONS A Note has been provided to modify the ACTIONS related to RCIC System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RCIC System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RCIC System instrumentation channel.

A.1 Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.23-1. The applicable Condition referenced in the Table is Function dependent.

Each time a channel is discovered to be inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.

B.1 and B.2 Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic initiation capability for the RCIC System. In (continued)

JAFNPP B 3.3.5.23-6 Revision XXX0

RCIC System Instrumentation B 3.3.5.23 BASES ACTIONS B.1 and B.2 (continued) this case, automatic initiation capability is lost if two Function 1 channels in the same trip system are inoperable and untripped. In this situation (loss of automatic initiation capability), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Action B.2 is not appropriate, and the RCIC System must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after discovery of loss of RCIC initiation capability.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."

For Required Action B.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically initiated due to two inoperable, untripped Reactor Vessel Water Level - Low Low (Level 2) channels in the same trip system. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been shown to be acceptable (Ref. 3) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.

Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition E must be entered and its Required Action taken.

C.1 A risk based analysis was performed and determined that an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Ref. 3) is acceptable to permit restoration of any inoperable channel to OPERABLE status (Required Action C.1). A Required Action (similar to Required Action B.1) limiting the allowable out (continued)

JAFNPP B 3.3.5.23-7 Revision 0XX

RCIC System Instrumentation B 3.3.5.23 BASES ACTIONS C.1 (continued) of service time, if a loss of automatic RCIC initiation capability exists, is not required. This Condition applies to the Reactor Vessel Water Level - High (Level 8) Function whose logic is arranged such that any inoperable channel will result in a loss of automatic RCIC initiation capability due to closure of the RCIC steam inlet valve. As stated above, this loss of automatic RCIC initiation capability was analyzed and determined to be acceptable.

This Condition also applies to the Manual Initiation Function. Since this Function is not assumed in any accident or transient analysis, a total loss of manual initiation capability (Required Action C.1) for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed. The Required Action does not allow placing a channel in trip since this action would not necessarily result in a safe state for the channel in all events.

D.1, D.2.1, and D.2.2 Required Action D.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in automatic component initiation capability being lost for the feature(s). For Required Action D.1, the RCIC System is the only associated feature. In this case, automatic initiation capability (automatic suction source alignment) is lost if two Function 3 channels associated with the same CST are inoperable and untripped. In this situation (loss of automatic suction source alignment), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Required Actions D.2.1 and D.2.2 is not appropriate, and the RCIC System must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of RCIC initiation capability. As noted, Required Action D.1 is only applicable if the RCIC pump suction is not aligned to the suppression pool since, if aligned, the Function is already performed.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."

For Required Action D.1, the Completion Time only begins upon discovery that the RCIC System suction source cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The (continued)

JAFNPP B 3.3.5.23-8 Revision 0XX

RCIC System Instrumentation B 3.3.5.23 BASES ACTIONS D.1, D.2.1, and D.2.2 (continued) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability (automatic suction source alignment) is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been shown to be acceptable (Ref. 3) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1, which performs the intended function of the channel (shifting the suction source to the suppression pool). Alternatively, Required Action D.2.2 allows the manual alignment of the RCIC suction to the suppression pool, which also performs the intended function. If Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the RCIC System piping remains filled with water. If it is not desired to perform Required Actions D.2.1 and D.2.2 (e.g., as in the case where shifting the suction source could drain down the RCIC suction piping), Condition E must be entered and its Required Action taken.

E.1 With any Required Action and associated Completion Time not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared inoperable immediately.

SURVEILLANCE As noted in the beginning of the SRs, the SRs for each RCIC REQUIREMENTS System instrumentation Function are found in the SRs column of Table 3.3.5.23-1.

The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed 'as follows:

(a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 2 and 4; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 1 and 3, provided the associated Function maintains trip capability. Upon completion of the (continued)

JAFNPP B 3.3.5.23-9 Revision 0XX

RCIC System Instrumentation B 3.3.5.23 BASES SURVEILLANCE Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be REQUIREMENTS returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 3) assumption of the average time required to perform channel surveillance.

That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the RCIC will initiate when necessary.

SR 3.3.5.23.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a parameter on other similar channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Channel agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

SR 3.3.5.23.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required (continued)

JAFNPP B 3.3.5.23-10 Revision 30XX

RPV Water Inventory Control Instrumentation B 3.3.5.23 BASES SURVEILLANCE SR 3.3.5.23.2 (continued)

REQUIREMENTS contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.5.23.3 and SR 3.3.5.23.5 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.

The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with the applicable extensions.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

JAFNPP B 3.3.5.23-11 Revision 30XX

RCIC System Instrumentation B 3.3.5.23 BASES SURVEILLANCE SR 3.3.5.23.2 REQUIREMENTS (continued) The calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.5.23-1. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.5.23.6 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to provide complete testing of the safety function.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50.36(c)(2)(ii).

2. Drawing 11825-5.01-15D, Rev. D, Reactor Assembly Nuclear Boiler, (GE Drawing 919D690BD).

3. GENE-770-06-2-A, Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, December 1992.

JAFNPP B 3.3.5.23-12 Revision 30XX

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE associated with these signals are addressed in LCO 3.3.5.1, SAFETY ANALYSES, "Emergency Core Cooling Systems (ECCS) Instrumentation," and LCO, and LCO 3.3.5.23, "Reactor Core Isolation Cooling (RCIC) System APPLICABILITY Instrumentation," and are not included in this LCO.

(continued)

In general, the individual Functions are required to be OPERABLE in MODES 1, 2, and 3 consistent with the Applicability for LCO 3.6.1.1, "Primary Containment."

Functions that have different Applicabilities are discussed below in the individual Functions discussion.

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.

Main Steam Line Isolation 1.a. Reactor Vessel Water Level - Low Low Low (Level 1)

Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.

Therefore, isolation of the MSIVs and other interfaces with the reactor vessel occurs to prevent offsite dose limits from being exceeded. The Reactor Vessel Water Level - Low Low Low (Level 1) Function is one of the many Functions assumed to be OPERABLE and capable of providing isolation signa1s. The Reactor Vessel Water Level - Low Low Low (Level 1) Function associated with isolation is assumed in the analysis of the recirculation line break (Ref. 2). The isolation of the MSLs on Level 1 supports actions to ensure that offsite dose limits are not exceeded for a DBA.

Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level - Low Low Low (Level 1) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level - Low Low Low (Level 1)

Allowable Value is chosen to ensure that the MSLs isolate on a potential loss of coolant accident (LOCA) to prevent offsite doses from exceeding 10 CFR 100 limits. In addition, the setting is low enough to allow the removal of heat from the reactor for a predetermined time following a (continued)

JAFNPP B 3.3.6.1-7 Revision 0XX

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 6.a. Reactor Pressure - High (continued)

SAFETY ANALYSES, LCO, and Each transmitter senses reactor pressure and provides input to each APPLICABILITY trip system. However, only one channel input is required to be OPERABLE for a trip system to be considered OPERABLE. Two channels of Reactor Pressure - High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Function is only required to be OPERABLE in MODES 1, 2, and 3, since these are the only MODES in which the reactor can be pressurized; thus, equipment protection is needed.

The Allowable Value was chosen to be low enough to protect the system equipment from overpressurization.

This Function isolates both RHR shutdown cooling pump suction valves.

6.b. Reactor Vessel Water Level - Low (Level 3)

Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of some reactor vessel interfaces occurs to begin isolating the potential sources of a break. The Reactor Vessel Water Level - Low (Level 3) Function associated with RHR Shutdown Cooling System isolation is not directly assumed in safety analyses because a break of the RHR Shutdown Cooling System is bounded by breaks of the reactor water recirculation system and MSL.

The RHR Shutdown Cooling System isolation on Level 3 supports actions to ensure that the RPV water level does not drop below the top of the active fuel during a vessel draindown event caused by a leak (e.g., pipe break or inadvertent valve opening) in the RHR Shutdown Cooling System.

Reactor Vessel Water Level - Low (Level 3) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels (two channels per trip system) of the Reactor Vessel Water Level - Low (Level 3) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. As noted (footnote (e) to Table 3.3.6.1-l), only one trip system of the Reactor Vessel Water Level - Low (Level 3)

Function are required to (continued)

JAFNPP B 3.3.6.1-22 Revision 27XX

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 6.b. Reactor Vessel Water Level - Low (Level 3) (continued)

SAFETY ANALYSES, LCO, and be OPERABLE in MODES 4 and 5, provided the RHR Shutdown APPLICABILITY Cooling System integrity is maintained. System integrity is maintained provided the piping is intact and no maintenance or other activity is being performed that has the potential for draining the reactor vessel through the system.

The Reactor Vessel Water Level - Low (Level 3) Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level-Low (Level 3) Allowable Value (LCO 3.3.1.1),

since the capability to cool the fuel may be threatened.

The Allowable Value is referenced from a level of water 352.56 inches above the lowest point in the inside bottom of the RPV and also corresponds to the top of a 144 inch fuel column (Ref. 13).

The Reactor Vessel Water Level - Low (Level 3) Function is only required to be OPERABLE in MODES 3, 4, and 5 to prevent this potential flow path from lowering the reactor vessel level to the top of the fuel. In MODES 1 and 2, another isolation (i.e., Reactor Pressure - High) and administrative controls ensure that this flow path remains isolated to prevent unexpected loss of inventory via this flow path.

This Function isolates both RHR shutdown cooling pump suction valves and the inboard LPCI injection valves.

Traversing Incore Probe System Isolation 7.a. Reactor Vessel Water Level-Low (Level 3)

Low RPV water level indicates that the capability to cool the fuel may be threatened. The valves whose penetrations communicate with the primary containment are isolated to limit the release of fission products. The isolation of the primary containment on Level 3 supports actions to ensure that offsite dose limits of 10 CFR 100 are not exceeded.

The Reactor Vessel Water Level-Low (Level 3) Function associated with isolation is implicitly assumed in the UFSAR analysis as these leakage paths are assumed to be isolated post LOCA.

Reactor Vessel Water Level - Low (Level 3) signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Two channels of Reactor (continued)

JAFNPP B 3.3.6.1-23 Revision XXX3

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS H.1 and H.2 (continued)

If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, or any Required Action of Condition F or G is not met and the associated Completion Time has expired, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

I.1 and I.2 If the channel is not restored to OPERABLE status within the allowed Completion Time, the associated SLC subsystem is declared inoperable or the RWCU System is isolated. Since this Function is required to ensure that the SLC System performs its intended function, sufficient remedial measures are provided by declaring the associated SLC subsystems inoperable or isolating the RWCU System.

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because it minimizes risk while allowing sufficient time for personnel to isolate the RWCU System.

J.1 and J.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated penetration flow path should be closed. However, if the shutdown cooling function is needed to provide core cooling, these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the channel to OPERABLE status or to isolate the RHR Shutdown Cooling System (i.e., provide alternate decay heat removal capabilities so the penetration flow path can be isolated). Actions must continue until the channel is restored to OPERABLE status or the RHR Shutdown Cooling System is isolated.

SURVEILLANCE As noted (Note 1) at the beginning of the SRs, the SRs for REQUIREMENTS each Primary Containment Isolation instrumentation Function are found in the SRs column of Table 3.3.6.1-1.

(continued)

JAFNPP B 3.3.6.1-29 Revision 3XX

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 1. Reactor Vessel Water Level - Low (Level 3) (continued)

SAFETY ANALYSES, LCO, and The Reactor Vessel Water Level - Low (Level 3) Allowable APPLICABILITY Value was chosen to be the same as the RPS level scram Allowable Value (LCO 3.3.1.1, "Reactor Protection System Instrumentation"), since this could indicate that the capability to cool the fuel is being threatened. The Allowable Value is referenced from a level of water 352.56 inches above the lowest point in the inside bottom of the RPV and also corresponds to the top of a 144 inch fuel column (Ref. 8).

The Reactor Vessel Water Level - Low (Level 3) Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, this Function is not required. In addition, the Function is also required to be OPERABLE during operations with a potential for draining the reactor vessel (OPDRVs) because the capability of isolating potential sources of leakage must be provided to ensure that offsite and control room dose limits are not exceeded if core damage occurs.

2. Drywell Pressure-High High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite and control room release. The Drywell Pressure - High Function is one of the Functions assumed to be OPERABLE and capable of providing isolation and initiating signals. The isolation and initiation systems on high drywell pressure supports actions to ensure that any offsite and control room releases are within the limits calculated in the safety analysis (Ref. 4).

High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure - High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the isolation function.

(continued)

JAFNPP B 3.3.6.2-4 Revision 0XX

CREVAS System Instrumentation B 3.3.7.1 BASES LCO personnel; thus, an alarm is provided in the control room so (continued) that the CREVAS System can be placed in the isolate mode of operation.

APPLICABILITY The Control Room Air Inlet Radiation - High Function is required to be OPERABLE in MODES 1, 2, and 3 and during OPDRVs and movement of recently irradiated fuel assemblies in the secondary containment, to ensure that control room personnel are protected during a LOCA, or fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g.,OPDRVs), the probability of a LOCA is low: thus, the Function is not required. Also due to radioactive decay, the Function is only required to provide an alarm to alert the operator of the need to initiate the CREVAS System during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ).

ACTIONS A.1 and A.2 With the Control Room Air Inlet Radiation - High Function inoperable one CREVAS subsystem must be placed in the isolate mode of operation per Required Action A.1 to ensure that control room personnel will be protected in the event of a Design Basis Accident. Alternately, if it is not desired to start a CREVAS subsystem, the CREVAS System must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is intended to allow the operator time to place the CREVAS subsystem in operation. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because it minimizes risk while allowing time for restoration of the channel, for placing one CREVAS subsystem in operation, or for entering the applicable Conditions and Required Actions for two inoperable CREVAS subsystems.

SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the low (continued)

JAFNPP B 3.3.7.1-3 Revision 2XX

ECCS Operating B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS Operating BASES BACKGROUND The ECCS is designed, in conjunction with the primary and secondary containment, to limit the release of radioactive materials to the environment following a loss of coolant accident (LOCA). The ECCS uses two independent methods (flooding and spraying) to cool the core during a LOCA. The ECCS network consists of the High Pressure Coolant Injection (HPCI) System, the Core Spray (CS) System, the low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System, and the Automatic Depressurization System (ADS). The suppression pool provides the required source of water for the ECCS.

Although no credit is taken in the safety analyses for the condensate storage tanks (CSTs), they are capable of providing a source of water for the HPCI and CS systems.

On receipt of an initiation signal, ECCS pumps automatically start; simultaneously, the system aligns and the pumps inject water, taken either from the CSTs or suppression pool, into the Reactor Coolant System (RCS) as RCS pressure is overcome by the discharge pressure of the ECCS pumps. Although the system is initiated, ADS action is delayed, allowing the operator to interrupt the timed sequence if the system is not needed. The HPCI pump discharge pressure almost immediately exceeds that of the RCS, and the pump injects coolant into the vessel to cool the core. If the break is small, the HPCI System will maintain coolant inventory as well as vessel level while the RCS is still pressurized. If HPCI fails, it is backed up by ADS in combination with LPCI and CS. In this event, if the ADS timed sequence is allowed to time out, the selected safety/relief valves (S/RVs) would open, depressurizing the RCS, thus allowing the LPCI and CS to overcome RCS pressure and inject coolant into the vessel. If the break is large, RCS pressure initially drops rapidly and the LPCI and CS cool the core.

Water from the break returns to the suppression pool where it is used again and again. Water in the suppression pool is circulated through a heat exchanger cooled by the RHR Service Water System.

Depending on the location and size of the break, portions of the (continued)

JAFNPP B 3.5.1-1 Revision 35XX

ECCS Operating B 3.5.1 BASES LCO realigned (remote or local) to the LPCI mode and not otherwise (continued) inoperable. Alignment and operation for decay heat removal includes when the system is being realigned from or to the RHR shutdown cooling mode. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling when necessary.

The HPCI system is considered OPERABLE when it is aligned to one or both CSTs with power available to support automatic realignment to the suppression pool if required. This is based on the design of the CSTs and the accident analysis which credits the suppression pool for supplying the HPCI System.

APPLICABILITY All ECCS subsystems are required to be OPERABLE during MODES 1, 2, and 3, when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system piping. In MODES 2 and 3, when reactor steam dome pressure is < 150 psig, ADS and HPCI are not required to be OPERABLE because the low pressure ECCS subsystems can provide sufficient flow below this pressure. ECCS rRequirements for MODES 4 and 5 are specified in LCO 3.5.2, "ECCS ShutdownRPV Water Inventory Control."

ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable HPCI System. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable HPCI System and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

A.1 If any one low pressure ECCS injection/spray subsystem is inoperable or if one LPCI pump in both LPCI subsystems is inoperable, the inoperable subsystem(s) must be restored to OPERABLE status within (continued)

JAFNPP B 3.5.1-6 Revision 35XX

RPV Water Inventory Control B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control BASES BACKGROUND The RPV contains penetrations below the top of the active fuel (TAF) that have the potential to drain the reactor coolant inventory to below the TAF.

If the water level should drop below the TAF, the ability to remove decay heat is reduced, which could lead to elevated cladding temperatures and clad perforation. Safety Limit 2.1.1.3 requires the RPV water level to be above the top of the active irradiated fuel at all times to prevent such elevated cladding temperatures.

APPLICABLE With the unit in MODE 4 or 5, RPV water inventory control is not SAFETY required to mitigate any events or accidents evaluated in the safety ANALYSES analyses. RPV water inventory control is required in MODES 4 and 5 to protect Safety Limit 2.1.1.3 and the fuel cladding barrier to prevent the release of radioactive material to the environment should an unexpected draining event occur.

A double-ended guillotine break of the Reactor Coolant System (RCS) is not postulated in MODES 4 and 5 due to the reduced RCS pressure, reduced piping stresses, and ductile piping systems. Instead, an event is considered in which single operator error or initiating event allows draining of the RPV water inventory through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e.g.,

seismic event, loss of normal power, single human error). It is assumed, based on engineering judgment, that while in MODES 4 and 5, one low pressure ECCS injection/spray subsystem can maintain adequate reactor vessel water level.

As discussed in References 1, 2, 3, 4, and 5, operating experience has shown RPV water inventory to be significant to public health and safety.

Therefore, RPV Water Inventory Control satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO The RPV water level must be controlled in MODES 4 and 5 to ensure that if an unexpected draining event should occur, the reactor coolant water level remains above the top of the active irradiated fuel as required by Safety Limit 2.1.1.3.

JAFNPP B 3.5.2-1 Revision XXX

RPV Water Inventory Control B 3.5.2 BASES LCO (continued)

The Limiting Condition for Operation (LCO) requires the DRAIN TIME of RPV water inventory to the TAF to be 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . A DRAIN TIME of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is considered reasonable to identify and initiate action to mitigate unexpected draining of reactor coolant. An event that could cause loss of RPV water inventory and result in the RPV water level reaching the TAF in greater than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months does not represent a significant challenge to Safety Limit 2.1.1.3 and can be managed as part of normal plant operation.

One low pressure ECCS injection/spray subsystem is required to be OPERABLE and capable of being manually started to provide defense-in-depth should an unexpected draining event occur. A low pressure ECCS injection/spray subsystem consists of either one Core Spray (CS) subsystem or one Low Pressure Coolant Injection (LPCI) subsystem.

Each CS subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or condensate storage tank (CST) to the RPV. Each LPCI subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool to the RPV. In MODES 4 and 5, the RHR System cross tie valve is not required to be closed.

The LCO is modified by a Note which allows a required LPCI subsystem to be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned (remote or local) to the LPCI mode and is not otherwise inoperable. Alignment and operation for decay heat removal includes when the required RHR pump is not operating or when the system is realigned from or to the RHR shutdown cooling mode. This allowance is necessary since the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Because of the restrictions on DRAIN TIME, sufficient time will be available following an unexpected draining event to manually align and initiate LPCI subsystem operation to maintain RPV water inventory prior to the RPV water level reaching the TAF.

APPLICABILITY RPV water inventory control is required in MODES 4 and 5.

Requirements on water inventory control in other MODES are contained in LCOs in Section 3.3, Instrumentation, and other LCOs in Section 3.5, ECCS, RCIC, and RPV Water Inventory Control. RPV water inventory control is required to protect Safety Limit 2.1.1.3 which is applicable whenever irradiated fuel is in the reactor vessel.

JAFNPP B 3.5.2-2 Revision XXX

RPV Water Inventory Control B 3.5.2 BASES ACTIONS A.1 and B.1 If the required low pressure ECCS injection/spray subsystem is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . In this Condition, the LCO controls on DRAIN TIME minimize the possibility that an unexpected draining event could necessitate the use of the ECCS injection/spray subsystem, however the defense-in-depth provided by the ECCS injection/spray subsystem is lost. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time for restoring the required low pressure ECCS injection/spray subsystem to OPERABLE status is based on engineering judgment that considers the LCO controls on DRAIN TIME and the low probability of an unexpected draining event that would result in loss of RPV water inventory.

If the inoperable ECCS injection/spray subsystem is not restored to OPERABLE status within the required Completion Time, action must be initiated immediately to establish a method of water injection capable of operating without offsite electrical power. The method of water injection includes the necessary instrumentation and controls, water sources, and pumps and valves needed to add water to the RPV or refueling cavity should an unexpected draining event occur. The method of water injection may be manually initiated and may consist of one or more systems or subsystems, and must be able to access water inventory capable of maintaining the RPV water level above the TAF for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

If recirculation of injected water would occur, it may be credited in determining the necessary water volume.

C.1, C.2, and C.3 With the DRAIN TIME less than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months but greater than or equal to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , compensatory measures should be taken to ensure the ability to implement mitigating actions should an unexpected draining event occur.

Should a draining event lower the reactor coolant level to below the TAF, there is potential for damage to the reactor fuel cladding and release of radioactive material. Additional actions are taken to ensure that radioactive material will be contained, diluted, and processed prior to being released to the environment.

The secondary containment provides a controlled volume in which fission products can be contained, diluted, and processed prior to release to the environment. Required Action C.1 requires verification of the capability to establish the [secondary] containment boundary in less than the DRAIN TIME. The required verification confirms actions to establish the secondary containment boundary are preplanned and necessary materials are available. The [secondary] containment boundary is considered established when one Standby Gas Treatment (SGT) subsystem is capable of maintaining a negative pressure in the

[secondary] containment with respect to the environment.

JAFNPP B 3.5.2-3 Revision XXX

RPV Water Inventory Control B 3.5.2 BASES ACTIONS (continued)

Verification that the secondary containment boundary can be established must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The required verification is an administrative activity and does not require manipulation or testing of equipment. Secondary containment penetration flow paths form a part of the secondary containment boundary. Required Action C.2 requires verification of the capability to isolate each secondary containment penetration flow path in less than the DRAIN TIME. The required verification confirms actions to isolate the secondary containment penetration flow paths are preplanned and necessary materials are available. Power operated valves are not required to receive automatic isolation signals if they can be closed manually within the required time.

Verification that the secondary containment penetration flow paths can be isolated must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The required verification is an administrative activity and does not require manipulation or testing of equipment.

One SGT subsystem is capable of maintaining the secondary containment at a negative pressure with respect to the environment and filter gaseous releases. Required Action C.3 requires verification of the capability to place one SGT subsystem in operation in less than the DRAIN TIME. The required verification confirms actions to place a SGT subsystem in operation are preplanned and necessary materials are available. Verification that a SGT subsystem can be placed in operation must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The required verification is an administrative activity and does not require manipulation or testing of equipment.

D.1, D.2, D.3, and D.4 With the DRAIN TIME less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , mitigating actions are implemented in case an unexpected draining event should occur. Note that if the DRAIN TIME is less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Required Action E.1 is also applicable.

Required Action D.1 requires immediate action to establish an additional method of water injection augmenting the ECCS injection/spray subsystem required by the LCO. The additional method of water injection includes the necessary instrumentation and controls, water sources, and pumps and valves needed to add water to the RPV or refueling cavity should an unexpected draining event occur. The Note to Required Action D.1 states that either the ECCS injection/spray subsystem or the additional method of water injection must be capable of operating without offsite electrical power. The additional method of water injection may be manually initiated and may consist of one or more systems or subsystems. The additional method of water injection must be able to JAFNPP B 3.5.2-4 Revision XXX

RPV Water Inventory Control B 3.5.2 BASES ACTIONS (continued) access water inventory capable of being injected to maintain the RPV water level above the TAF for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The additional method of water injection and the ECCS injection/spray subsystem may share all or part of the same water sources. If recirculation of injected water would occur, it may be credited in determining the required water volume.

Should a draining event lower the reactor coolant level to below the TAF, there is potential for damage to the reactor fuel cladding and release of radioactive material. Additional actions are taken to ensure that radioactive material will be contained, diluted, and processed prior to being released to the environment.

The secondary containment provides a control volume in which fission products can be contained, diluted, and processed prior to release to the environment. Required Action D.2 requires that actions be immediately initiated to establish the secondary containment boundary. With the secondary containment boundary established, one SGT subsystem is capable of maintaining a negative pressure in the secondary containment with respect to the environment.

The secondary containment penetrations form a part of the secondary containment boundary. Required Action D.3 requires that actions be immediately initiated to verify that each secondary containment penetration flow path is isolated or to verify that it can be manually isolated from the control room.

One SGT subsystem is capable of maintaining the secondary containment at a negative pressure with respect to the environment and filter gaseous releases. Required Action D.4 requires that actions be immediately initiated to verify that at least one SGT subsystem is capable of being placed in operation. The required verification is an administrative activity and does not require manipulation or testing of equipment.

E.1 If the Required Actions and associated Completion times of Conditions C or D are not met or if the DRAIN TIME is less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , actions must be initiated immediately to restore the DRAIN TIME to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . In this condition, there may be insufficient time to respond to an unexpected draining event to prevent the RPV water inventory from reaching the TAF.

Note that Required Actions D.1, D.2, D.3, and D.4 are also applicable when DRAIN TIME is less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

JAFNPP B 3.5.2-5 Revision XXX

RPV Water Inventory Control B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 REQUIREMENTS This Surveillance verifies that the DRAIN TIME of RPV water inventory to the TAF is 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The period of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is considered reasonable to identify and initiate action to mitigate draining of reactor coolant. Loss of RPV water inventory that would result in the RPV water level reaching the TAF in greater than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months does not represent a significant challenge to Safety Limit 2.1.1.3 and can be managed as part of normal plant operation.

The definition of DRAIN TIME states that realistic cross-sectional areas and drain rates are used in the calculation. A realistic drain rate may be determined using a single, step-wise, or integrated calculation considering the changing RPV water level during a draining event. For a Control Rod RPV penetration flow path with the Control Rod Drive Mechanism removed and not replaced with a blank flange, the realistic cross-sectional area is based on the control rod blade seated in the control rod guide tube. If the control rod blade will be raised from the penetration to adjust or verify seating of the blade, the exposed cross-sectional area of the RPV penetration flow path is used.

The definition of DRAIN TIME excludes from the calculation those penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank flanges, or other devices that prevent flow of reactor coolant through the penetration flow paths. A blank flange or other bolted device must be connected with a sufficient number of bolts to prevent draining in the event of an Operating Basis Earthquake. Normal or expected leakage from closed systems or past isolation devices is permitted. Determination that a system is intact and closed or isolated must consider the status of branch lines and ongoing plant maintenance and testing activities.

JAFNPP B 3.5.2-6 Revision XXX

RPV Water Inventory Control B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)

The Residual Heat Removal (RHR) Shutdown Cooling System is only considered an intact closed system when misalignment issues (Reference 6) have been precluded by functional valve interlocks or by isolation devices, such that redirection of RPV water out of an RHR subsystem is precluded. Further, RHR Shutdown Cooling System is only considered an intact closed system if its controls have not been transferred to Remote Shutdown, which disables the interlocks and isolation signals.

The exclusion of penetration flow paths from the determination of DRAIN TIME must consider the potential effects of a single operator error or initiating event on items supporting maintenance and testing (rigging, scaffolding, temporary shielding, piping plugs, snubber removal, freeze seals, etc.). If failure of such items could result and would cause a draining event from a closed system or between the RPV and the isolation device, the penetration flow path may not be excluded from the DRAIN TIME calculation.

Surveillance Requirement 3.0.1 requires SRs to be met between performances. Therefore, any changes in plant conditions that would change the DRAIN TIME requires that a new DRAIN TIME be determined.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.5.2.2 and SR 3.5.2.3 The minimum water level of 10.33 ft required for the suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the CS subsystem or LPCI subsystem pump, recirculation volume, and vortex prevention.

With the suppression pool water level less than the required limit, the JAFNPP B 3.5.2-7 Revision XXX

TSTF-542, Rev. 2 RPV Water Inventory Control B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued) required ECCS injection/spray subsystem is inoperable unless aligned to an OPERABLE CST.

The required CS System is OPERABLE only if it can take suction from the CST, and the CST water level is sufficient to provide the required NPSH for the CS pump. Therefore, a verification that either the suppression pool water level is 10.33 ft or that a required CS subsystem is aligned to take suction from the CST and the CST contains 354,000 gallons (two tanks) of water, equivalent to 324 inches (27 ft),

ensures that the CS subsystem can supply at least 50,000 gallons of makeup water to the RPV. The CS suction is uncovered at the 258,000 gallon (two tanks) level.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.5.2.4 The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the required ECCS injection/spray subsystems full of water ensures that the ECCS subsystem will perform properly. This may also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is to vent at the high points.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.5.2.5 Verifying the correct alignment for manual, power operated, and automatic valves in the required ECCS subsystem flow path provides assurance that the proper flow paths will be available for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time.

This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

JAFNPP B 3.5.2-8 Revision XXX

RPV Water Inventory Control B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.2.6 Verifying that the required ECCS injection/spray subsystem can be manually started and operate for at least 10 minutes demonstrates that the subsystem is available to mitigate a draining event. Testing the ECCS injection/spray subsystem through the recirculation line is necessary to avoid overfilling the refueling cavity. The minimum operating time of 10 minutes was based on engineering judgement.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.5.2.7 Verifying that each valve credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated RPV water level isolation signal is required to prevent RPV water inventory from dropping below the TAF should an unexpected draining event occur.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.5.2.8 The required ECCS injection/spray subsystem shall be capable of being manually operated from the Control Room. This Surveillance verifies that the required CS or LPCI subsystem (including the associated pump and valve(s)) can be manually operated to provide additional RPV water inventory, if needed.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.

JAFNPP B 3.5.2-9 Revision XXX

RPV Water Inventory Control B 3.5.2 BASES REFERENCES 1. Information Notice 84-81 "Inadvertent Reduction in Primary Coolant Inventory in Boiling Water Reactors During Shutdown and Startup,"

November 1984.

2. Information Notice 86-74, "Reduction of Reactor Coolant Inventory Because of Misalignment of RHR Valves," August 1986.

3. Generic Letter 92-04, "Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f), " August 1992.

4. NRC Bulletin 93-03, "Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs," May 1993.

5. Information Notice 94-52, "Inadvertent Containment Spray and Reactor Vessel Draindown at Millstone 1," July 1994.

6. General Electric Service Information Letter No. 388, "RHR Valve Misalignment During Shutdown Cooling Operation for BWR 3/4/5/6,"

February 1983.

JFNPP B 3.5.2-10 Revision XXX

RCIC System B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.3 RCIC System BASES BACKGROUND The RCIC System is not part of the ECCS; however, the RCIC System is included with the ECCS section because of their similar functions.

The RCIC System is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of the RPV water level.

Under these conditions, the High Pressure Coolant Injection (HPCI) and RCIC systems perform similar functions. The RCIC System design requirements ensure that the criteria of Reference 1 are satisfied.

The RCIC System (Ref. 2) consists of a steam driven turbine pump unit, piping and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger. Suction piping is provided from the condensate storage tanks (CSTs) and the suppression pool. Pump suction is normally aligned to the CSTs to minimize injection of suppression pool water into the RPV. However, if the CST water supply is low, an automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System. The steam supply to the turbine is piped from the "B" main steam line upstream of the associated inboard main steam line isolation valve.

The RCIC System is designed to provide core cooling for a wide range of reactor pressures (150 psig to 1195 psig). Upon receipt of an initiation signal, the RCIC turbine accelerates to a specified speed.

As the RCIC flow increases, the turbine control valve is automatically adjusted to maintain design flow. Exhaust steam from the RCIC turbine is discharged to the suppression pool. A full flow test line is provided to route water to the CSTs to allow testing of the RCIC System during normal operation without injecting water into the RPV.

(continued)

JAFNPP B 3.5.3-1 Revision 29XXX

RCIC System B 3.5.3 BASES BACKGROUND The RCIC pump is provided with a minimum flow bypass line, which (continued) discharges to the suppression pool. The valve in this line automatically opens to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects, the RCIC System discharge piping is kept full of water. The RCIC System is normally aligned to the CSTs. The height of water in the CSTs is sufficient to maintain the piping full of water up to the first isolation valve. The relative height of the feedwater line connection for RCIC is such that the water in the feedwater lines keeps the remaining portion of the RCIC discharge line full of water.

Therefore, RCIC does not require a "keep full" system.

APPLICABLE SAFETY The function of the RCIC System is to respond to transient ANALYSES events by providing makeup coolant to the reactor. The RCIC System is not an Engineered Safeguard System and no credit is taken in the safety analyses for RCIC System operation. The RCIC System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii) (Ref. 3).

LCO The OPERABILITY of the RCIC System provides adequate core cooling such that actuation of any of the low pressure ECCS subsystems is not required in the event of RPV isolation accompanied by a loss of feedwater flow. The RCIC System has sufficient capacity for maintaining RPV inventory during an isolation event.

The RCIC system is considered OPERABLE when it is aligned to one or both CSTs with power available to support automatic realignment to the suppression pool if required. This is based on the design of the CSTs and the RCIC System.

APPLICABILITY The RCIC System is required to be OPERABLE during MODE 1, and MODES 2 and 3 with reactor steam dome pressure > 150 psig, since RCIC is the primary non-ECCS water source for core cooling when the reactor is isolated and pressurized. In MODES 2 and 3 with reactor steam dome pressure <150 psig, the low pressure ECCS injection/spray subsystem can provide sufficient flow to the RPV. and iIn MODES 4 and 5, RCIC is not required to be OPERABLE since RPV water inventory control is required by LCO 3.5.2, RPV Water Level Inventory Control.the low pressure ECCS injection/spray subsystems can provide sufficient flow to the RPV.

(continued)

JAFNPP B 3.5.3-2 Revision XXX29

PCIVs B 3.6.1.3 BASES LCO flanges are in place, and closed systems are intact. These (continued) passive isolation valves and devices are those listed in Reference 8.

MSIVs, Low Pressure Coolant Injection (LPCI) and Core Spray (CS) System air operated testable check valves must meet additional leakage rate requirements. Other PCIV leakage rates are addressed by LCO 3.6.1.1. "Primary Containment."

as Type B or C testing.

This LCO provides assurance that the PCIVs will perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the primary containment boundary during accidents.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, most PCIVs are not required to be OPERABLE and the primary containment vent and purge valves are not required to be normally closed in MODES 4 and 5.

Certain valves, however, are required to be OPERABLE when the to prevent inadvertent reactor vessel draindown. These valves are those whose associated instrumentation is required to be OPERABLE per LCO 3.3.6.1. "Primary Containment Isolation Instrumentation." (This does not include the valves that isolate the associated instrumentation.)

ACTIONS The ACTIONS are modified by a Note allowing penetration flow path(s) to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator at the controls of the valve. who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.

A second Note has been added to provide clarification that.

for the purpose of this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable PCIV.

Complying with the Required Actions may allow for continued operation, and subsequent inoperable PCIVs are governed by subsequent Condition entry and application of associated Required Actions.

(continued)

JAFNPP B 3.6.1.3-4 Revision 0XX

PCIVs B 3.6.1.3 BASES ACTIONS G.1 and G.2 (continued) suspend operations with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended and valve(s) are restored to OPERABLE status. If suspending an OPDRV would result in closing the residual heat removal (RHR) shutdown cooling isolation valves, an alternative Required Action is provided to immediately initiate action to restore the valve(s) to OPERABLE status. This allows RHR shutdown cooling to remain in service while actions are being taken to restore the valve.

SURVEILLANCE SR 3.6.1.3.1 REQUIRMENTS This SR ensures that the primary containment vent and purge valves are closed as required or, if open, open for an allowable reason. If a purge valve is open in violation of this SR, the valve is considered inoperable. The SR is modified by a Note stating that the SR is not required to be met when the vent and purge valves are open for the stated reasons. The Note states that these valves may be opened for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open, provided the full-flow 12 inch line (with valve 27MOV-120) to the SGT System is closed and one or more SGT System reactor building suction valves are open. This will ensure there is no damage to the filters if a LOCA were to occur with the vent and purge valves open since excessive differential pressure is not expected with the full-flow 12 inch line closed and one or more SGT System reactor building suction valves open. The 20 and 24 inch vent and purge valves are capable of closing against the dynamic effects of a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.6.1.3.2 This SR ensures that each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed or otherwise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits.

(continued)

JAFNPP B 3.6.1.3-10 Revision 30XXX

Suppression Pool Water Level B 3.6.2.2 BASES APPLICABLE loads due to S/RV discharges. Suppression pool water level SAFETY ANALYSES must be maintained within the limits specified so that the (continued) safety analysis of References 1 and 2 remain valid.

Suppression pool water level satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 3).

LCO A limit that suppression pool water level be > 13.88 ft and

< 14 ft is required to ensure that the primary containment conditions assumed for the safety analyses are met. Either the high or low water level limits were used in the safety analyses, depending upon which is more conservative for a particular calculation.

The LCO is modified by a note which states that the LCO is not required to be met up to four hours during Surveillances that cause suppression pool water level to be outside of limits. These Surveillances include required OPERABILITY testing of the High Pressure Coolant Injection System, the Reactor Core Isolation Cooling System, the suppression chamber-to-drywell vacuum breakers, the Core Spray System and the Residual Heat Removal System. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowance is adequate to perform the Surveillances and to restore the suppression pool water level to within limits.

APPLICABILITY In MODES 1, 2, and 3, a DBA would cause significant loads on the primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. The requirement for maintaining suppression pool water level within limits in MODE 4 or 5 is addressed in LCO 3.5.2.

Reactor Pressure Vessel (RPV) Water Inventory ControlECCS -

Shutdown.

ACTIONS A.1 With suppression pool water level outside the limits, the conditions assumed for the safety analyses are not met. If water level is below the minimum level, the pressure suppression function still exists as long as the vent system downcomer lines are covered, HPCI and RCIC turbine exhausts are covered, and S/RV quenchers are covered. If suppression pool water level is above the maximum level, protection against overpressurization still exists due to the margin in the peak containment pressure analysis and the capability of the Residual Heat Removal Containment Spray System.

(continued)

JAFNPP B 3.6.2.2-2 Revision 0XXX

Secondary Containment B 3.6.4.1 BASES APPLICABLE Secondary containment satisfies Criterion 3 of SAFETY ANALYSES 10 CFR 50.36(c)(2)(ii) (Ref. 3).

(continued)

LCO An OPERABLE secondary containment provides a control volume into which fission products that leak from primary containment, or are released from the reactor coolant pressure boundary components located in secondary containment, or are released directly to the secondary containment as a result of a refueling accident, can be processed prior to release to the environment. For the secondary containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained.

APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.

In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay, secondary containment is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e.,

fuel that has occupied part of a critical reactor core within the previous 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ).

ACTIONS A.1 If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.

(continued)

JAFNPP B 3.6.4.1-2 Revision 2XXX

Secondary Containment B 3.6.4.1 BASES ACTIONS B.1 and B.2 (continued)

If secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 and C.2 Movement of recently irradiated fuel assemblies in the secondary containment and OPDRVs can be postulated to cause significant fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the environment. Therefore, movement of recently irradiated fuel assemblies must be immediately suspended if the secondary containment is inoperable.

Suspension of this activity shall not preclude completing an action that involves moving a component to a safe position.

Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

LCO 3.0.3 is not applicable in MODES 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5.

LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.

Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.1.1 REQUIREMENTS This SR ensures that the secondary containment boundary is sufficiently leak tight to preclude exfiltration under expected wind conditions. Momentary transients on the (continued)

JAFNPP B 3.6.4.1-3 Revision 2XXX

SCIVs B 3.6.4.2 BASES APPLICABLE Maintaining SCIVs OPERABLE with isolation times within SAFETY ANALYSES limits ensures that fission products will remain trapped (continued) inside secondary containment so that they can be treated by the SGT System prior to discharge to the environment.

SCIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii)

(Ref. 3).

LCO SCIVs form a part of the secondary containment boundary.

The SCIV safety function is related to control of offsite radiation releases resulting from DBAs.

The power operated automatic isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves covered by this LCO, along with their associated stroke times, are listed in Reference 4.

The normally closed isolation valves or blind flanges are considered OPERABLE when manual valves are closed or open in accordance with appropriate administrative controls, automatic SCIVs are de-activated and secured in their closed position, and blind flanges are in place. These passive isolation valves or devices are listed in Reference 4.

APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore, the OPERABILITY of SCIVs is required.

In MODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for situations under which significant radioactive releases can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. Moving recently irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and

3. Due to radioactivitveradioactive decay, SCIVs are only requriedrequired to be OPERABLE during fuel handling involving recently irradiated fuel (i.e ., fuel that has occupied part of a critical reactor core within the previous 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ).

(continued)

JAFNPP B 3.6.4.2-2 Revision XXX2

SCIVs B 3.6.4.2 BASES ACTIONS C.1 and C.2 (continued)

If any Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

D.1 and D.2 If any Required Action and associated Completion Time are not met, the plant must be placed in a condition in which the LCO does not apply. If applicable, the movement of recently irradiated fuel assemblies in the secondary containment must be immediately suspended. Suspension of this activity shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action D.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving fuel while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.2.1 REQUIREMENTS This SR verifies that each secondary containment manual isolation valve and blind flange that is not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the secondary containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification that those (continued)

JAFNPP B 3.6.4.2-5 Revision XXX2

SGT System B 3.6.4.3 BASES (continued)

APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, SGT System OPERABILITY is required during these MODES.

In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT System in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay, the SGT system is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ).

ACTIONS A.1 With one SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status in 7 days. In this Condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT subsystem and the low probability of a DBA occurring during this period.

B.1 and B.2 If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable.

based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

JAFNPP B 3.6.4.3-3 Revision 2XXX

SGT System B 3.6.4.3 BASES ACTIONS C.1, and C.2.1, and C.2.2 (continued)

During movement of recently irradiated fuel assemblies, in the secondary containment or during OPDRVs, when Required Action A.1 cannot be completed within the required Completion Time, the OPERABLE SGT subsystem should immediately be placed in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected.

An alternative to Required Action C.1 is to immediately suspend activities that represent a potential for releasing a significant amount of radioactive material to the secondary containment, thus placing the plant in a condition that minimizes risk. If applicable, movement of recently irradiated fuel assemblies must immediately be suspended.

Suspension of this activity must not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

LCO 3.0.3 is not applicable in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the Required Actions of Condition C have been modified by a Note stating that LCO 3.0.3 is not applicable.

If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

D.1 If both SGT subsystems are inoperable in MODE 1, 2, or 3, the SGT System may not be capable of supporting the required radioactivity release control function. Therefore, action is required to enter LCO 3.0.3 immediately.

(continued)

JAFNPP B 3.6.4.3-4 Revision 2XXX

SGT System B 3.6.4.3 BASES ACTIONS E.1 and E.2 (continued)

When two SGT subsystems are inoperable, if applicable, movement of recently irradiated fuel assemblies in secondary containment must immediately be suspended. Suspension of this activity shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

LCO 3.0.3 is not applicable in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action E.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating each SGT subsystem fan for > 10 continuous hours ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for > 10 continuous hours periodically eliminates moisture on the adsorbers and HEPA filters. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.6.4.3.2 This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP.

(continued)

JAFNPP B 3.6.4.3-5 Revision 30XXX

CREVAS System B 3.7.3 BASES LCO b. A prefilter, two HEPA filters and charcoal adsorbers are not (continued) excessively restricting flow and are capable of performing their filtration functions; and

c. Ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

In order for the CREVAS subsystems to be considered OPERABLE, the CRE boundary must be maintained such that the CRE occupant dose from a large radioactive release does not exceed the calculated dose in the licensing basis consequence analyses for DBAs, and that CRE occupants are protected from hazardous chemicals and smoke.

The LCO is modified by a Note allowing the CRE boundary to be opened intermittently under administrative controls. This Note only applies to openings in the CRE boundary that can be rapidly restored to the design condition, such as doors, hatches, floor plugs, and access panels. For entry and exit through doors the administrative control of the opening is performed by the person{s) entering or exiting the area. For other openings, these controls should be proceduralized and consist of stationing a dedicated individual at the opening who is in continuous communication with the operators in the CRE. This individual will have a method to rapidly close the opening and to restore the CRE boundary to a condition equivalent to the design condition when a need for CRE isolation is indicated.

APPLICABILITY In MODES 1, 2, and 3, the CREVAS System must be OPERABLE to ensure that the CRE will remain habitable during and following a DBA, since the DBA could lead to a fission product release.

In MODES 4 and 5, the probability and consequences of a DBA are reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaining the CREVAS System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a. During operations with a potential for draining the reactor vessel (OPDRVs);and

b. Dduring movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay, the CREVAS system is only requriedrequired to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor within the previous 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ).

(continued)

JAFNPP B 3.7.3-3 Revision 20XXX

CREVAS System B 3.7.3 BASES ACTIONS C.1 and C.2 (continued)

In MODE 1, 2, or 3, if the inoperable CREVAS subsystem or CRE boundary cannot be restored to OPERABLE status within the required Completion Time, the plant must be placed in a MODE that minimizes accident risk. To achieve this status, the plant must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

D.1. D.2.1. and D.2.2 LCO 3.0.3 is not applicable when in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the Required Actions of Condition D are modified by a Note indicating that LCO 3.0.3 does not apply.

If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs, if the inoperable CREVAS subsystem cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE CREVAS subsystem may be placed in the isolate mode. This action ensures that the remaining subsystem is OPERABLE, and that any active failure will be readily detected.

An alternative to Required Action D.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the CRE. This places the plant in a condition that minimizes the accident risk.

If applicable, movement of recently irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of this activity shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Action must continue until the OPDRVs are suspended.

(continued)

JAFNPP B 3.7.3-5 Revision 20XXX

CREAVS System B 3.7.3 BASES ACTIONS E.1 (continued)

If both CREVAS subsystems are inoperable in MODE 1, 2, or 3 for reasons other than an inoperable CRE boundary (i.e., Condition B),

the CREVAS System may not be capable of performing the intended function and the plant is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.

F.1 and F.2 LCO 3.0.3 is not applicable when in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODES 1, 2, or 3, the Required Actions of Condition F are modified by a Note indicating that LCO 3.0.3 does not apply. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs, with two CREVAS subsystems inoperable or with one or more CREVAS subsystems inoperable due to an inoperable CRE boundary, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room.

This places the plant in a condition that minimizes the accident risk.

If applicable, movement of recently irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of this activity shall not preclude completion of movement of a component to a safe position. If applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until the OPDRVs are suspended.

SURVEILLANCE SR 3.7.3.1 REQUIREMENTS This SR verifies that a subsystem in a standby mode starts on demand and continues to operate. These subsystems should be checked periodically to ensure that they start and function properly.

As the environmental and normal operating conditions of this system are not severe, testing each subsystem once every three months provides an adequate check on this system. Since the CREVAS System does not contain heaters, it need only be operated for

> 15 minutes to demonstrate the function of the system. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

JAFNPP B 3.7.3-6 Revision 30XXX

Control Room AC System B 3.7.4 BASES APPLICABLE component failure of a component of the Control Room AC System, SAFETY ANALYSES assuming a loss of offsite power, does not impair the ability of the (continued) system to perform its design function. Redundant detectors and controls are provided for control room temperature control. The Control Room AC System is designed in accordance with Seismic Category I requirements. The Control Room AC System is capable of removing sensible and latent heat loads from the control room, including consideration of equipment heat loads and personnel occupancy requirements to ensure equipment OPERABILITY.

The Control Room AC System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 2).

LCO Two redundant subsystems of the Control Room AC System are required to be OPERABLE to ensure that at least one is available, assuming a single active component failure disables the other subsystem. Total system failure could result in the equipment operating temperature exceeding limits.

The Control Room AC System is considered OPERABLE when the individual components necessary to maintain the control room temperature are OPERABLE in both subsystems. These components include the air handling units, recirculation exhaust fans, air handling unit fans, ductwork, dampers, and associated Instrumentation and controls. The cooling calls of the air handling units may be cooled by the control room chillers, but to satisfy this LCO the Emergency Service Water System must be capable of alignment to provide cooling water directly to the cooling coils.

APPLICABILITY In MODE 1, 2, or 3, the Control Room AC System must be OPERABLE to ensure that the control room temperature will not exceed equipment OPERABILITY limits following control room isolation.

In MODES 4 and 5, the probability and consequences of a Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Control Room AC System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a. During operations with a potential for draining the reactor vessel (OPDRVs); and

b. Dduring movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay, the Control Room AC system is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ).

(continued)

JAFNPP B 3.7.4-2 Revision 20XXX

Control Room AC System B 3.7.4 BASES (continues)

ACTIONS A.1 With one control room AC subsystem inoperable, the inoperable control room AC subsystem must be restored to OPERABLE status within 30 days. With the plant in this condition, the remaining OPERABLE control room AC subsystem is adequate to perform the control room air conditioning function. However, the overall reliability is reduced because a single active component failure in the OPERABLE subsystem Could result in loss of the control room air conditioning function. The 30 day Completion Time is based on the low probability of an event occurring requiring control room isolation, the consideration that the remaining subsystem can provide the required protection, and the availability of alternate safety and nonsafely cooling methods.

B.1. and B.2 If both control room AC subsystems are inoperable, the Control Room AC System may not be capable of performing its intended function.

Therefore, the control room area temperature is required to be monitored to ensure that temperature is being maintained low enough that equipment in the control room is not adversely affected. With the control room temperature being maintained within the temperature limit, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to restore a control room AC subsystem to OPERABLE status. This Completion Time is reasonable considering that the control room temperature is being maintained within limits and the low probability of an event occurring requiring control room isolation.

C.1 and C.2 In MODE 1, 2, or 3, if the inoperable control room AC subsystem cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed In a MODE that minimizes risk. To achieve this status, the plant must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

D.1, D.2.1 and D.2.2 LCO 3.0.3 is not applicable while in MODE 4 and 5. However, since recently irradiated fuel assembly movement can occur in MODES 1, 2, or 3 the Required Actions of condition D are modified by a Note indicating that LCO 3.0.3 does not apply.

(continued)

JAFNPP B 3.7.4-3 Revision 20

Control Room AC System B 3.7.4 BASES ACTIONS D.1, D.2.1 and D.2.2 (continued)

If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs, if Required Action A.1 cannot be completed within the required Completion Time, the OPERABLE control room AC subsystem may be placed immediately in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent actuation will occur, and that any active failure will be readily detected.

An alternative to Required Action D.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the plant in a condition that minimizes risk.

If applicable, movement of recently irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of this activity shall not preclude completion of movement of a component to a safe position. Also, If applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until the OPDRVs are suspended.

E.1 and E.2 LCO 3.0.3 is not applicable when in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3 the Required Actions of Condition E are modified by a Note indicating that LCO 3.0.3 does not apply.

If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown.

During movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs, if Required Actions B.1 and B.2 cannot be met within the required Completion Times, action must be taken to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the plant in a condition that minimizes risk.

(continued)

JAFNPP B 3.7.4-4 Revision 20XXX

Control Room AC System B 3.7.4 BASES ACTIONS E.1 and E.2 (continued)

If applicable, handling of recently irradiated fuel in the secondary containment must be suspended immediately. Suspension of this activity shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until the OPDRVs are suspended.

SURVEILLANCE SR 3.7.4.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the control room heat load assumed in the safety analyses with ESW providing water to the cooling coils of the air handling units. Heat transfer testing is not performed on the Control Room (CR) and Relay Room (RR) Air Handlin Units (AHUs) as these coolers are closed loop, glycol based systems which are not prone to fouling. To verify the system has the capability to remove the assumed heat, the ESW supply function (safety related) is required to be operable and the following surveillance requirements met: 1) the manual valves needed to initiate ESW flow to these coolers are cycled to verify operability; 2) the ESW supply piping to the AHUs is flushed during check valve testing; and 3) flow rates are measured against target flow rates. Therefore, any degradation would be detected and corrected through the corrective action program. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program JAF calculations verify maximum Allowable Tube Plugging Limit for CR and RR AHUs if maintenance is required on the AHUs. The level of allowed plugging provides a margin in the CR and RR equipment heat load and still maintains the CR and RR below 104° F under accident conditions using ESW at 85° F. In addition, JAF calculations state the potential for plugged tubes is low crediting use of a closed loop cooling water system using glycol/demineralized water (not service water) as the cooling medium.

REFERECES 1. UFSAR, Section 9.9.3.11.

2. 10 CFR 50.36(c)(2)(ii).

3. SEP-SW-001 Rev.0, NRC Generic Letter 89-13 Service Water Program JAFNPP B 3.7.4-5 Revision 30

AC Sources - Shutdown B 3.8.2 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources - Shutdown BASES BACKROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC Sources-Operating." In addition to the reserve AC sources described in LCO 3.8.1, during plant shutdown with the main generator off line, the plant emergency buses may be supplied using the 345 kV (backfeed)

AC source. The 345 kV backfeed requires removing the main generator disconnect links that tie the main generator to the 24 kV bus, and providing power from the 345 kV transmission network to energize the main transformers (T1A and T1B), 24 kV bus, normal station service transformer (NSST) 71T-4, and subsequent 4.16 kV distribution and emergency buses. The 345 kV offsite backfeed AC source as well as the two (2) 115 kV offsite circuits are the qualified offsite circuits during outages.

APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 SAFETY ANALYSES and 5 and during movement of recently irradiated fuel assemblies in the secondary containment ensures that:

a. The facility can be maintained in the shutdown or refueling condition for extended periods:

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the plant status:

and

c. Adequate AC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident involving handling recently irradiated fuel.

Due to radioactive decay, AC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ).

In general, when the plant is shutdown the Technical Specifications requirements ensure that the plant has the capability to mitigate the consequences of postulated accidents. However, assuming a single active component failure and concurrent loss of all offsite or loss of all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, and 3 have no specific (continued)

JAFNPP B 3.8.2-1 Revision 7XX

AC Sources - Shutdown B 3.8.2 BASES (continued)

LCO One qualified offsite circuit capable of supplying one division of the plant Class IE AC power distribution subsystem(s) of LCO 3.8.8, "Distribution Systems -Shutdown,"

and one qualified offsite circuit, which may be the same circuit required above, capable of supplying the other division of the plant Class IE AC power distribution subsystem(s) when a second division is required by LCO 3.8.8, ensures that all required loads are powered from offsite power. An OPERABLE EDG subsystem, associated with a 4.16 kV emergency bus required OPERABLE by LCO 3.8.8, ensures that a diverse power source is available for providing electrical power support assuming a loss of the offsite circuit. Together, OPERABILITY of the required offsite circuit and EDG subsystem ensures the availability of sufficient AC sources to operate the plant in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents involving handling recently irradiated fuel and reactor vessel draindown). Automatic initiation of the required EDG during shutdown conditions is specified in LCO 3.3.5.1, "ECCS Instrumentation," and LCO 3.3.8.1, "LOP Instrumentation."

The qualified offsite circuit(s) must be capable of maintaining rated frequency and voltage while connected to its respective 4.16 kV emergency bus(es), and of accepting required loads during an accident. Qualified offsite circuits are those that are described in LCO 3.8.1 Bases and the UFSAR and are part of the licensing basis for the plant.

However, since the plant is shutdown, when two offsite circuits are required, they may share one of the incoming switchyard breakers provided the North and South bus disconnect is closed. Also, while in this condition, the automatic opening feature of the disconnect is not required to be OPERABLE. This is allowed since the two offsite circuits are not required to be independent while shutdown.

The required EDG subsystem must be capable of starting, accelerating to rated speed and voltage, force paralleling, and connecting to its respective emergency bus on detection of bus undervoltage. This sequence must be accomplished within 11 seconds. The required EDG subsystem must also be capable of accepting required loads within the assumed loading sequence intervals, and must continue to operate until offsite power can be restored to the emergency buses.

These capabilities are required to be met with the EDG subsystem in standby condition.

(continued)

JAFNPP B 3.8.2-3 Revision XXX2

AC Sources - Shutdown B 3.8.2 BASES (continued)

LCO Proper sequencing of loads. including tripping of (continued) nonessential loads. is a required function for EDG subsystem OPERABILITY. The necessary portions of the Emergency Service Water System and Ultimate Heat Sink are also required to provide appropriate cooling to the required EDG subsystem. In addition. proper sequence operation is an integral part of offsite circuit OPERABILITY since its inoperability impacts the ability to start and maintain energized loads required OPERABLE by LCO 3.8.8.

No automatic transfer capability is required for offsite circuits to be considered OPERABLE.

APPLICABILITY The AC sources are required to be OPERABLE in MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the secondary containment to provide assurance that:

a. Systems that provide core cooling providing adequate coolant inventory makeup are available for the irradiated fuel assemblies in

a. the core in case of an inadvertent draindown of the reactor vessel;

b. Systems needed to mitigate a fuel handling accident involving handling recently irradiated fuel (i.e.,

fuel that has occupied part of a critical reactor core within the previous 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ) are available;

c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the plant in a cold shutdown condition or refueling condition.

AC power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.1.

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However.

since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5.

LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.

Entering LCO 3.0.3. while in MODE 1. 2. or 3 would require the unit to be shutdown unnecessarily.

(continued)

JAFNPP B 3.8.2-4 Revision XXX2

AC Sources - Shutdown B 3.8.2 JAFNPP B 3.8.2-4 Revision XXX2

AC Sources - Shutdown B 3.8.2 BASES ACTIONS A.1 (continued)

An offsite circuit is considered inoperable if it is not available to one required 4.16 kV emergency bus. If two 4.16 kV emergency buses are required per LCO 3.8.8, one division with offsite power available may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, and recently irradiated fuel movement, and operations with a potential for draining the reactor vessel. By the allowance of the option to declare required features inoperable with no offsite power, appropriate restrictions can be implemented in accordance with the affected required feature(s) LCOs' ACTIONS. These required features are those that are assumed in the safety analysis to function to mitigate events postulated during shutdown, such as an inadvertent draindown of the reactor vessel or a fuel handling accident involving recently irradiated fuel. These required features do not include monitoring requirements.

A.2.1, A.2.2, A.2.3, A.2.4, B.1, B.2, and B.3, and B.4 With an offsite circuit not available to all required 4.16 kV emergency buses, the option still exists to declare all required features inoperable per Required Action A.1. Since this option may involve undesired administrative efforts, the allowance for sufficiently conservative actions is made. With the required EDG subsystem inoperable, the minimum required diversity of AC power sources is not available. It is, therefore, required to suspend CORE ALTERATIONS and, movement of recently irradiated fuel assemblies in the secondary containment., and activities that could result in inadvertent draining of the reactor vessel.

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC sources and to continue this action until restoration is accomplished in order to provide the necessary AC power to the plant safety systems.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required AC electrical power sources should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.

Pursuant to LCO 3.0.6, the Distribution System ACTIONS would not be entered even if all AC sources to it are inoperable, resulting in de-energization.

Therefore, the Required Actions of Condition A have been modified by a Note to indicate that when Condition A is entered with no AC power to any (continued)

JAFNPP B 3.8.2-5 Revision XXX14

AC Sources - Shutdown B 3.8.2 BASES ACTIONS A.2.1, A.2.2, A.2.3, A.2.4, B.1, B.2, and B.3, and B.4 (continued) required 4.16 kV emergency bus, ACTIONS for LCO 3.8.8 must be immediately entered. This Note allows Condition A to provide requirements for the loss of an offsite circuit whether or not a division is de-energized. LCO 3.8.8 provides the appropriate restrictions for the situation involving a de-energized division.

SURVEILLANCE SR 3.8.2.1 REQUIREMENTS SR 3.8.2.1 requires the SRs from LCO 3.8.1 that are necessary for ensuring the OPERABILITY of the AC sources in other than MODES 1, 2, and 3. SR 3.8.1.7 is not required to be met since the main generator is not used to provide AC power while shutdown. Refer to the corresponding Bases for LCO 3.8.1 for a discussion of each SR.

This SR is modified by two Notes. The reason for Note 1 is to preclude requiring the OPERABLE EDG subsystem from being paralleled with the reserve power network or otherwise rendered inoperable during the performance of SRs, and to preclude de-energizing a required 4.16 kV emergency bus or disconnecting a required reserve circuit during performance of SRs. With limited AC sources available, a single event could compromise both the required reserve circuit and EDG subsystem. It is the intent that these SRs must still be capable of being met, but actual performance is not required during periods when the EDG subsystem and reserve circuit is required to be OPERABLE.

Note 2 states that SRs 3.8.1.10 and 3.8.1.12 are not required to be met when its associated ECCS subsystem(s) are not required to be OPERABLE.

These SRs demonstrate the EDG response to an ECCS signal (either alone or in conjunction with a loss of power signal). This is consistent with the ECCS instrumentation requirements that do not require the ECCS signal when the ECCS System is not required to be OPERABLE per LCO 3.5.2, "ECCS-Shutdown."

REFERENCES 1. 10 CFR 50.36(c)(2)(ii).

JAFNPP B 3.8.2-6 Revision 14XX

DC Sources - Shutdown B 3.8.5 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC Sources - Shutdown BASES BACKROUND A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC Sources - Operating."

APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the UFSAR, Chapter 6 (Ref. 1) and Chapter 14 (Ref. 2), assume that Engineered Safeguards systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the emergency diesel generators (EDGs), emergency auxiliaries, and control and switching during all MODES of operation and during movement of recently irradiated fuel assemblies in the secondary containment.

The OPERABILITY of the DC subsystems is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum DC electrical power sources during MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the secondary containment ensures that:

a. The facility can be maintained in the shutdown or refueling condition for extended periods:

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the plant status: and

c. Adequate DC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a refueling accident involving handling recently irradiated fuel.

Due to radioactive decay, DC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ).

In general, when the unit is shutdown, the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and (continued)

JAFNPP B 3.8.5-1 Revision 2XXX

DC Sources - Shutdown B 3.8.5 BASES APPLICABLE concurrent loss of all offsite or all onsite power is not SAFETY ANALYSES required. The rationale for this is based on the fact that (continued) many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, and 3 have no specific analyses in MODES 4 and 5. Worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and the corresponding stresses result in the probabilities of occurrence being significantly reduced or eliminated, and in minimal consequences. These deviations from DBA analysis assumptions and design requirements during shutdown conditions are allowed by the LCO for required systems.

The shutdown Technical Specification requirements are designed to ensure that the unit has the capability to mitigate the consequences of certain postulated accidents.

Worst case Design Basis Accidents which are analyzed for operating MODES are generally viewed not to be a significant concern during shutdown MODES due to the lower energies involved. The Technical Specifications therefore require a lesser complement of electrical equipment to be available during shutdown than is required during operating MODES.

More recent work completed on the potential risks associated with shutdown, however, have found significant risk associated with certain shutdown evolutions. As a result, in addition to the requirements established in the Technical Specifications, the industry has adopted NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management," as an Industry initiative to manage shutdown tasks and associated electrical support to maintain risk at an acceptable low level. This may require the availability of additional equipment beyond that required by the shutdown Technical Specifications.

The DC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii)

(Ref. 3).

LCO One 125 VDC electrical power subsystem consisting of one 125 V battery, one battery charger, and the corresponding control equipment and interconnecting cabling supplying power to the associated bus is required to be OPERABLE to support one DC distribution subsystem required OPERABLE by LCO 3.8.8, "Distribution Systems - Shutdown." This requirement ensures the availability of sufficient DC electrical power sources to operate the plant in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., refueling accidents involving handling recently irradiated fuel and inadvertent reactor vessel draindown).

(continued)

JAFNPP B 3.8.5-2 Revision 2XXX

DC Sources - Shutdown B 3.8.5 BASES APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the secondary containment provide assurance that:

a. Required features to provide core cooling adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel;

b. Required features needed to mitigate a fuel handling accident involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ) are available;

c. Required features necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the plant in a cold shutdown condition or refueling condition.

The DC electrical power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.4.

ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2 or 3, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Entering LCO 3.0.3, while in MODE 1, 2, or 3 would require the unit to be shutdown unnecessarily.

A.1, A.2.1, A.2.2, and A.2.3, and A.2.4 By allowance of the option to declare required features inoperable with the associated DC electrical power subsystem inoperable, appropriate restrictions are implemented in accordance with the affected system LCOs' ACTIONS. These required features are those that are assumed in the safety analysis to function to mitigate events postulated during shutdown, such as an inadvertent draindown of the reactor vessel or a fuel handling accident involving recently irradiated fuel. These required features do not include monitoring requirements.

However, in many instances this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS, movement of recently irradiated fuel assemblies in the (continued)

JAFNPP B 3.8.5-3 Revision 14XXX

DC Sources - Shutdown B 3.8.5 BASES ACTIONS A.1, A.2.1, A.2.2, and A.2.3 and A.2.4 (continued) secondary containment, and any activities that could result in inadvertent draining of the reactor vessel). Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required DC electrical power subsystem and to continue this action until restoration is accomplished in order to provide the necessary DC electrical power to the plant safety systems.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required DC electrical power subsystems should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.

SURVEILLANCE SR 3.8.5.1 REQUIREMENTS SR 3.8.5.1 requires performance of all Surveillances required by SR 3.8.4.1 through SR 3.8.4.4. Therefore, see the corresponding Bases for LCO 3.8.4 for a discussion of each SR.

This SR is modified by a Note. The reason for the Note is to preclude requiring the OPERABLE DC electrical power subsystem from being discharged below their capability to provide the required power supply or otherwise rendered inoperable during the performance of SRs. It is the intent that these SRs must still be capable of being met, but actual performance is not required.

REFERENCES 1. UFSAR, Chapter 6.

2. UFSAR, Chapter 14.

3. 10 CFR 50.36(c)(2)(ii).

JAFNPP B 3.8.5-4 Revision 14XXX

Distribution Systems - Shutdown B 3.8.8 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Distribution Systems - Shutdown BASES BACKROUND A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC Sources - Operating."

APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the UFSAR. Chapter 6 (Ref. 1) and Chapter 14 (Ref. 2), assume Engineered Safeguards systems are OPERABLE. The AC and 125 VDC electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to Engineered Safeguards systems so that the fuel. Reactor Coolant System. And containment design limits are not exceeded.

The OPERABILITY of the AC and 125 VDC electrical power distribution systems is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.

The OPERABILITY of the minimum AC and 125 VDC electrical power sources and associated power distribution subsystems during MODES 4 and 5. and during movement of recently irradiated fuel assemblies in the secondary containment ensures that:

a. The facility can be maintained in the shutdown or refueling condition for extended periods:

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the plant status: and

c. Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a refueling accident involving handling recently irradiated fuel. Due to radioactive decay, DC electrical power is only required to mitigate fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ).

The AC and 125 VDC electrical power distribution systems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 3).

(continued)

JAFNPP B 3.8.8-1 Revision 2XXX

Distribution Systems - Shutdown B 3.8.8 BASES (continued)

LCO Various combinations of subsystems, equipment, and components are required OPERABLE by other LCOs, depending on the specific plant condition. Implicit in those requirements is the required OPERABILITY of necessary support required features. This LCO explicitly requires energization of the portions of the electrical distribution system necessary to support OPERABILITY of Technical Specification required systems, equipment, and components - both specifically addressed by their own LCO, and implicitly required by the definition of OPERABILITY.

Maintaining these portions of the distribution system energized ensures the availability of sufficient power to operate the plant in a safe manner to mitigate the consequences of postulated events during shutdown (e.g.,

fuel handling accidents involving handling recently irradiated fuel and inadvertent reactor vessel draindown).

APPLICABILITY The AC and 125 VDC electrical power distribution subsystems required to be OPERABLE in MODES 4 and 5 and during movement of recently irradiated fuel assemblies in the secondary containment provide assurance that:

a. Systems that provide core cooling to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel;

b. Systems needed to mitigate a fuel handling accident involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ) are available;

c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the plant in a cold shutdown condition or refueling condition.

The AC, and 125 VDC electrical power distribution subsystem requirements for MODES 1, 2, and 3 are covered in LCO 3.8.7.

(continued)

JAFNPP B 3.8.8-2 Revision 2XXX

Distribution Systems - Shutdown B 3.8.8 BASES ACTIONS LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Entering LCO 3.0.3, while in MODE 1, 2, or 3 would require the unit to be shutdown unnecessarily.

A.1, A.2.1, A.2.2, A.2.3, and A.2.4, and A.2.5 Although redundant required features may 'require redundant divisions of electrical power distribution subsystems to be OPERABLE, one OPERABLE distribution subsystem division may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS and, recently irradiated fuel movement., and operations with a potential for draining the reactor vessel. By allowing the option to declare required features associated with an inoperable distribution subsystem inoperable, appropriate restrictions are implemented in accordance with the affected distribution subsystem LCO's Required Actions. These required features are those that are assumed in the safety analysis to function to mitigate events postulated during shutdown, such as an inadvertent draindown of the reactor vessel or a fuel handling accident involving recently irradiated fuel. These required features do not include monitoring requirements. In many instances this option may involve undesired administrative efforts.

Therefore, the allowance for sufficiently conservative actions is made, (i.e., to suspend CORE ALTERATIONS and, movement of recently irradiated fuel assemblies in the secondary containment, and any activities that could result in inadvertent draining of the reactor vessel).

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and 125 VDC electrical power distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the plant safety systems.

Not Withstanding performance of the above conservative Required Actions, a required residual heat removal-shutdown cooling (RHR-SDC) subsystem may be inoperable. In this case, Required Actions A.2.1 through A.2.4 do not adequately address the concerns relating to coolant circulation and heat removal. Pursuant to LCO 3.0.6, the RHR-SDC ACTIONS would not be entered. Therefore, Required Action A.2.5 is provided to direct declaring RHR-SDC inoperable, which results in taking the appropriate RHR-SDC ACTIONS.

(continued)

JAFNPP B 3.8.8-3 Revision 14XXX

Inservice Leak and Hydrostatic Test Operation B 3.10.1 BASES APPLICABLE during the performance of hydrostatic or leak testing. The SAFETY ANALYSES required pressure testing conditions provide adequate (continued) assurance that the consequences of a recirculation line break (Refs. 2 and 3) will be conservatively bounded by the consequences of the postulated main steam line break outside of primary containment described in Reference 4. Therefore, these requirements will conservatively limit radiation releases to the environment.

In the unlikely event of a largeany primary system leak that could result in draining of the RPV, the reactor vessel would rapidly depressurize, allowing the low pressure core cooling systems to operate. The makeup capability of the low pressure coolant injection and core spray subsystems, as required in MODE 4 by LCO 3.5.2. "ECCS - ShutdownReactor Pressure Vessel (RPV) Water Inventory Control." would be more than adequate to keep the RPV water level above the TAF core flooded under this low decay heat load condition.

Small system leaks would be detected by leakage inspections before significant inventory loss occurred.

For the purposes of this test, the protection provided by normally required MODE 4 applicable LCOs, in addition to the secondary containment requirements required to be met by this Special Operations LCO, will ensure acceptable consequences during normal hydrostatic test conditions and during postulated accident conditions.

As described in LCO 3.0.7. compliance with Special Operations LCOs is optional, and therefore, no criteria of 10 CFR 50.36(c)(2)(ii) (Ref. 5) apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.

LCO As described in LCO 3.0.7. compliance with this Special Operations LCO is optional. Operation at reactor coolant temperatures> 212°F can be in accordance with Table 1.1-1 for MODE 3 operation without meeting this Special Operations LCO or its ACTIONS. This option may be required due to P/T limits, however, which require testing at temperatures

> 212°F. while performance of inservice leak and hydrostatic testing results in inoperability of subsystems required when

> 212°F.

(continued)

JAFNPP B 3.10.1-2 Revision 0XXX

v t e Letter Sequence Request
EPID:L-2017-LLA-0311, TSTF-542 (Approved, Closed)
Initiation Request Acceptance... Supplement Administration Questions 23 March 2018 26 March 2018 Answers 19 April 2018 Results Approval Other:
MONTHYEAR2018-03-23 [Table View]

JAFP-17-0093, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2: Difference between revisions

Revision as of 16:41, 29 October 2019

Text

Subject:

Subject:

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

4.0 ENVIRONMENTAL CONSIDERATION

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JAFP-17-0093, Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2: Difference between revisions

Revision as of 16:41, 29 October 2019

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1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

4.0 ENVIRONMENTAL CONSIDERATION

5.0 REFERENCES

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

4.0 ENVIRONMENTAL CONSIDERATION

5.0 REFERENCES

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