JAFP-23-0040, License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis

License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis

ML23215A012
Person / Time
Site:	FitzPatrick
Issue date:	08/03/2023
From:	David Gudger Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
JAFP-23-0040
Download: ML23215A012 (1)

Text

200 Exelon Way Kennett Square, PA 19348 www.constellation.com 10 CFR 50.90 JAFP-23-0040 August 3, 2023 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 James A. FitzPatrick Nuclear Power Plant Renewed Facility Operating License No. DPR-59 NRC Docket No. 50-333

Subject:

License Amendment Request to Update the Technical Specification Bases to Change the Fuel Handling Accident Analysis In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) requests approval of proposed changes to the Technical Specification (TS) Bases to change the Fuel Handling Accident Analysis (FHA) due to new Refuel Bridge Mast NF-400 to NF-500 and definition for Recently Irradiated Fuel at James A. FitzPatrick Nuclear Power Plant (JAF). The impact of the proposed amendment would increase the consequences of the FHA analysis as well as reduce future refueling outage burden.

The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in of this submittal.

The proposed TS Bases markup pages are included as Attachment 2 to this submittal.

This amendment request contains no new regulatory commitments.

CEG requests approval of the proposed amendment by August 3, 2024. Once approved, the amendment shall be implemented within 45 days.

The proposed changes have been reviewed by the JAF Plant Operations Review Committee in accordance with the requirements of the Constellation Quality Assurance Program.

License Amendment Request Fuel Handling Accident Docket No. 50-333 August 3, 2023 Page 2 In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), CEG is transmitting a copy of this application and its attachments to the designated State Officials.

Should you have any questions concerning this submittal, please contact Abul Hasanat Abul.Hasanat@Constellation.com.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 3rd day of August 2023.

Respectfully, David T. Gudger Senior Manager - Licensing & Regulatory Affairs Constellation Energy Generation, LLC Attachments:

1. Evaluation of Proposed Changes

2. Proposed Technical Specification Bases Marked-Up Pages cc:

USNRC Region I, Regional Administrator w/attachments USNRC Senior Resident Inspector, JAF w/attachments USNRC Project Manager, JAF w/attachments A. L. Peterson, NYSERDA w/attachments

ATTACHMENT 1 License Amendment Request James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 EVALUATION OF PROPOSED CHANGES Subject License Amendment Request to Update Technical Specification Bases to Change the Fuel Handling Accident Analysis 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

License Amendment Request Fuel Handling Accident Page 1 of 6 Docket No. 50-333 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION Constellation Energy Generation, LLC (CEG) proposes to update the FitzPatrick Nuclear Power Plant (JAF) Technical Specifications (TS) Bases to Change the Fuel Handling Accident Analysis.

Specifically, CEG requests approval of proposed changes to JAF Alternative Source Term (AST) analysis of the Fuel Handling Accident (FHA).

The proposed amendment will revise the JAF FHA Analysis and TS Bases definition of Recently Irradiated Fuel to account for changes to the analyses in support of the transition from the Refuel Bridge Mast NF-400 (i.e., Triangular Mast) to the new NF-500 Mast. The impact of the proposed amendment would increase the consequences of the FHA analysis as well as reduce future refueling outage burden.

2.0 DETAILED DESCRIPTION The JAF FHA AST analysis was first implemented in 2002 after NRC approval with Amendment

1. 276 (ML022350228). This proposed amendment includes two major changes to the existing licensing basis, both of which would increase dose consequences sufficiently that prior NRC approval is required.

The first proposed change consists of creating a single unified radial peaking factor (RPF) for the new NF-500 Refueling Mast as well as all fuel types in operation at JAF, specifically GNF2 and GNF3 fuel.

The second proposed change involves a redefinition of recently irradiated fuel as it relates to Secondary Containment (SC) and Control Room Ventilation (CREVAS) operability. For SC operability, the definition of recently irradiated fuel will be changed from the current 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> down to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Modeling of SC operability requires automatic action, consistent with the current licensing basis. Overall, this change will allow for an earlier relaxation of SC operability to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown. This definition is applicable to:

TS Table 3.3.6.2, Secondary Containment Isolation Instrumentation, Items #3 & #4: RB Exhaust Radiation & Refueling Floor Exhaust Radiation TS 3.6.4.1, Secondary Containment TS 3.6.4.2, Secondary Containment Isolation Valves (SCIVs)

TS 3.6.4.3, Standby Gas Treatment (SGT) System For CREVAS operability, the definition of recently irradiated fuel will be changed from the current 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> up to 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br />. Modeling of CREVAS operability requires manual operator action as currently controlled by operator procedures. This change in modeling will require a longer requirement for CREVAS operability out to 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br /> after reactor shutdown. This definition is applicable to:

TS 3.3.7.1, CREVAS System Instrumentation TS 3.7.3, CREVAS System TS 3.7.4, Control Room Air Conditioning (AC) System TS 3.8.2, Electrical Power Systems/AC Sources Shutdown

License Amendment Request Fuel Handling Accident Page 2 of 6 Docket No. 50-333 Evaluation of Proposed Changes TS 3.8.5, Electrical Power Systems/DC Sources Shutdown TS 3.8.8, Electrical Power Systems/Distribution Systems-Shutdown These changes are in support of the planned utilization of the NF-500 mast for refueling starting with refueling outage J1R26 (September 2024). The current FHA analysis does not support the RPF with existing core loading strategies and the NF-500 mast. Therefore, the two major changes outlined above are introduced into the FHA analysis.

3.0 TECHNICAL EVALUATION

The JAF FHA AST analysis has undergone two previous changes by way of 10 CFR 50.59 Screening and Evaluation since the 2002 NRC approval of Amendment 276. The first change in 2018 utilized a higher RPF with the GNF2 fuel currently in use at that time. The second change in 2021 incorporated the new GNF3 fuel type. This proposed amendment includes two major changes to the existing licensing basis, both of which would increase dose consequences sufficiently that prior NRC approval is required.

The current analysis involves two cases which both use the applicable regulatory criteria of 5 rem Total Effective Dose Equivalent (TEDE) in the Control Room (CR) for the duration of the event, 6.3 rem TEDE at the exclusion area boundary (EAB) for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and 6.3 rem TEDE at the outer boundary of the low population zone (LPZ) for the duration of the event. Case 1 accounts for the scenario from 0 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after reactor shutdown with SC operable. Case 2 accounts for the scenario greater than 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after reactor shutdown where SC is no longer operable. The current dose results and associated applicable regulatory criteria are shown in Table 1:

Table 1 (JAF FHA Results - Current Licensing Basis) 30-day CR TEDE (rem)

Max 2-hour EAB TEDE (rem) 30-day LPZ TEDE (rem)

Case 1 1.95 2.72 0.304 Case 2 4.68 0.274 0.0306 Limit 5

6.3 6.3 The first proposed change consists of creating a single unified RPF of 1.70 for the new NF-500 Refueling Mast as well as all fuel types in operation at JAF, specifically GNF2 and GNF3 fuel.

The second proposed change involves a redefinition of recently irradiated fuel as it relates to SC and CREVAS operability. For SC operability, the definition of recently irradiated fuel will be changed from the current 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> down to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For CREVAS operability, the definition of recently irradiated fuel will be changed from the current 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> up to 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br />.

Implementation of these two proposed changes first required the FHA analysis to be redefined with three different cases. Case 1 accounts for the scenario from 0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown with SC operable. SC actuation occurs by 2-minute automatic actuation, consistent with the existing licensing basis. While CREVAS would be operable during Case 1, it is not credited in the analysis.

License Amendment Request Fuel Handling Accident Page 3 of 6 Docket No. 50-333 Evaluation of Proposed Changes Case 2 accounts for the scenario from 24 - 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br /> after reactor shutdown where SC is not operable, but CREVAS is operable. CREVAS action occurs by manual operator action as controlled by existing operator procedure AOP-44 Dropped Fuel Assembly. Crediting of CREVAS as early as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown requires an 8-minute Time Critical Action (TCA) after initiation of the event to isolate control room ventilation. The operator action is defined by the following criteria:

Procedure is entered when the control room operator in communication with the refuel floor reports a fuel handling accident.

Success criteria is achieved when the control room is isolated by manual switch in the control room and a supply isolation bypass damper is manually positioned closed to address single failure of inlet isolation valve.

This revision to the time critical operator action required time will be managed per guidance in OP-AA-102-106, Operator Response Time Program.

Case 3 accounts for the scenario greater than 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br /> after reactor shutdown where SC and CREVAS are no longer credited and therefore are not required to be operable. No automatic or manual actions are modeled within Case 3.

The final dose results for all three cases due to these changes and associated applicable regulatory criteria are shown in Table 2 for GNF2 fuel and Table 3 for GNF3 fuel:

Table 2 (JAF FHA Results - Proposed Changes for GNF2 Fuel) 30-day CR TEDE (rem)

Max 2-hour EAB TEDE (rem) 30-day LPZ TEDE (rem)

Case 1 2.13 2.98 0.333 Case 2 4.95 0.541 0.0605 Case 3 4.94 0.285 0.0319 Limit 5

6.3 6.3 Table 3 (JAF FHA Results - Proposed Changes for GNF3 Fuel) 30-day CR TEDE (rem)

Max 2-hour EAB TEDE (rem) 30-day LPZ TEDE (rem)

Case 1 2.03 2.84 0.317 Case 2 4.71 0.515 0.0576 Case 3 4.71 0.271 0.0303 Limit 5

6.3

6.3 CONCLUSION

S/FINDINGS The final results documented in Tables 2 & 3 remain within the applicable regulatory criteria established in Table 1 and would become the new limiting analysis of record for the JAF FHA analysis.

License Amendment Request Fuel Handling Accident Page 4 of 6 Docket No. 50-333 Evaluation of Proposed Changes

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The following regulatory requirements have been considered:

10 CFR 50.67, "Accident source term," establishes acceptable radiation dose limits resulting from design basis accidents for an individual located at the exclusion area boundary or low population zone, and for occupants of the control room. The analyses performed by Constellation demonstrate that the calculated radiological consequences of a design basis Refueling Accident (Fuel Handling Accident) meet the radiation dose limits specified in 10 CFR 50.67.

Regulatory Guide (RG) 1.183, dated July 2000, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," provides guidance for implementation of 10 CFR 50.67, including assumptions and methods that are acceptable to the NRC staff for performing design basis radiological analyses using an AST.

Regulatory Issue Summary (RIS) 2006-04, "Experience with Implementation of Alternative Source Terms," provides guidance to ensure that the appropriate level of technical detail is considered in AST analyses and included in AST submittals.

4.2 Precedent Letter from G. Vissing, Sr (NRC Project Manager) to M. Kansler (Sr. Vice President and Chief Operating Officer, Entergy Nuclear Operations, Inc.), "James A. FitzPatrick Nuclear Power Plant - Amendment Re: Technical Specification Change To The Requirements For Handling Irradiated Fuel Assemblies (Tac No. MB5328)" dated September 12, 2002 (ML022350228).

4.3 No Significant Hazards Consideration Constellation has evaluated the proposed change for the James A. FitzPatrick Nuclear Power Plant (JAF) and has determined that the proposed change does not involve a significant hazards consideration and is providing the responses to the following three questions to support a finding of no significant hazards consideration.

1.

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

Response: No The proposed amendment does not change any behavior or operation of the Fuel Handling Equipment due to transition from Refuel Bridge Mast NF-400 to NF-500 and time definition for Recently Irradiated Fuel.

License Amendment Request Fuel Handling Accident Page 5 of 6 Docket No. 50-333 Evaluation of Proposed Changes The proposed change results in an increase in the FHA radiological dose to a Control Room occupant. However, the resultant FHA Control Room dose consequences remain within the acceptance criteria provided by the NRC for use with the AST. These criteria are presented in 10 CFR 50.67 and Regulatory Guide 1.183. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

The proposed amendment does not result in a significant increase in the probability or consequences of any previously evaluated accident.

2.

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

Response: No The NF-500 mast is similar enough in design and function to the NF-400 mast to not create the possibility of a new or different kind of accident. The proposed change does not significantly alter the Fuel Handling system design, create new failure modes, or change any modes of operation. Consequently, there are no new initiators that could result in a new or different kind of accident.

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

3.

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

Response: No The change preserves the original design requirements and ensures adequate validation to confirm continued design capability. The margin of safety is considered to be that provided by meeting the applicable regulatory limits. The change to the FHA modeling results in an increase in Control Room dose following the FHA; however, since the Control Room dose following the design basis accident remains within the regulatory limits, there is not a significant reduction in a margin of safety.

Based on the above, CEG concludes that the proposed change to the Fuel Handling Accident analysis does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of no significant hazards consideration is justified.

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

License Amendment Request Fuel Handling Accident Page 6 of 6 Docket No. 50-333 Evaluation of Proposed Changes

5.0 ENVIRONMENTAL CONSIDERATION

Constellation has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation."

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 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review,"

paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. James A. FitzPatrick, Final Safety Analysis Report, Revision 28, April 2023.

2. Constellation Calculation JAF-CALC-RAD-04410, Fuel Handling Accident - AST Analysis for Relaxation of Secondary Containment Operability, Revision 4.

ATTACHMENT 2 License Amendment Request James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 PROPOSED TECHNICAL SPECIFICATION BASES MARKED-UP PAGES

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES JAFNPP B 3.3.6.2-5 Revision 3, 4. Reactor Building and Refueling Floor Ventilation Exhaust Radiation High (continued) actuation of the SGT System are initiated to limit the release of fission products as assumed in the UFSAR safety analyses (Refs. 4 and 5).

The Exhaust Radiation High signals are initiated from radiation detectors that are located on the ventilation exhaust piping coming from the reactor building and the refueling floor zones. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. Two channels of Reactor Building Ventilation Exhaust Radiation High Function and two channels of Refueling Floor Ventilation Exhaust Radiation High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Values are chosen to promptly detect gross failure of the fuel cladding and are set in accordance with the ODCM.

The Reactor Building and Refueling Floor Ventilation Exhaust Radiation High Functions are required to be OPERABLE in MODES 1, 2, and 3 where considerable RCS energy exists; 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, these Functions are not required. In addition, the Functions are also required to be OPERABLE during OPDRVs and movement of recently irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure that offsite and control room dose limits are not exceeded. Due to radioactive decay, the Function is only required to isolate secondary containment 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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

ACTIONS Note has been provided to modify the ACTIONS related to secondary containment isolation 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 APPLICABLE SAFETY ANALYSIS, LCO, and APPLICABILITY (continued)

CREVAS System Instrumentation B 3.3.7.1 BASES JAFNPP B 3.3.7.1-3 Revision APPLICABILITY The Control Room Air Inlet Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 and during 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. During MODES 4 and 5, 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 96 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br />).

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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is intended to allow the operator time to place the CREVAS subsystem in operation. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 REQUIREMENTS 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 for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 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 probability of an event requiring this Function during this time period and since many other alarms are available to indicate whether a design basis event has occurred.

SR 3.3.7.1.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK will detect (continued)

(continued)

Secondary Containment B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment BASES JAFNPP B 3.6.4.1-1 Revision BACKGROUND The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA). In conjunction with operation of the Standby Gas Treatment (SGT) System and closure of certain valves whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment.

The secondary containment is a structure that surrounds the primary containment and is designed to provide secondary containment for postulated loss-of-coolant accidents inside the primary containment.

The Secondary Containment also surrounds the refueling facilities and is designed to provide primary containment for the postulated refueling accident. This structure forms a control volume that serves to hold up and dilute the fission products. It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g., due to pump and motor heat load additions). To prevent ground level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure.

Requirements for these systems are specified separately in LCO 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT) System."

APPLICABLE SAFETY ANALYSIS There are two principal accidents for which credit is taken for secondary containment OPERABILITY. These are a loss of coolant accident (LOCA) (Ref. 1) and a refueling accident involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 96 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) inside secondary containment (Ref. 2). The secondary containment performs no active function in response to each of these limiting events; however, its leak tightness is required to ensure that fission products entrapped within the secondary containment structure will be treated by the SGT System prior to discharge to the environment.

(continued)

Secondary Containment B 3.6.4.1 BASES JAFNPP B 3.6.4.1-2 Revision Secondary containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii)

(Ref. 3).

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.

APLLICABILITY 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 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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

ACTIONS A.1 If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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.

APPLICABLE SAFETY ANALYSIS (continued)

(continued)

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

BASES JAFNPP B 3.6.4.2-1 Revision BACKGROUND The function of the SCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Refs. 1 and 2).

Secondary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that fission products that leak from primary containment following a DBA, or that are released during certain operations when primary containment is not required to be OPERABLE or take place outside primary containment, are maintained within the secondary containment boundary.

The OPERABILITY requirements for SCIVs help ensure that an adequate secondary containment boundary is maintained during and after an accident by minimizing potential paths to the environment.

These isolation devices consist of either passive devices or active (automatic) devices. Manual valves, de-activated automatic valves secured in their closed position (including check valves with flow through the valve secured), and blind flanges are considered passive devices.

Automatic SCIVs close on a secondary containment isolation signal to establish a boundary for untreated radioactive material within secondary containment following a DBA or other accidents.

Other penetrations are isolated by the use of valves in the closed position or blind flanges.

APPLICABLE SAFETY ANALYSIS The SCIVs must be OPERABLE to ensure the secondary containment barrier to fission product releases is established. The principal accidents for which the secondary containment boundary is required are a loss of coolant accident (Ref. 1) and a refueling accident involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 96 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) inside secondary containment (Ref. 2). The secondary containment performs no active function in response to either of these limiting events, but the boundary established by SCIVs is required to ensure that leakage from the primary containment is processed by the Standby Gas Treatment (SGT) System before being released to the (continued)

SCIVs B 3.6.4.2 BASES JAFNPP B 3.6.4.2-2 Revision environment.

Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped 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.

APLLICABILITY 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 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 radioactive decay, SCIVs are only required 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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

APPLICABLE SAFETY ANALYSIS (continued)

(continued)

SGT System B 3.6.4.3 BASES JAFNPP B 3.6.4.3-2 Revision The SGT System equipment and components are sized to reduce and maintain the secondary containment at a negative pressure of 0.25 inches water gauge when the system is in operation under neutral wind conditions and the SGT fans exhausting at a rate of 6,000 cfm.

The demister is provided to remove entrained water in the air, while the electric heater reduces the relative humidity of the airstream to less than 70% (Ref. 2). The prefilter removes large particulate matter, while the HEPA filter removes fine particulate matter and protects the charcoal from fouling. The charcoal adsorber removes gaseous elemental iodine and organic iodides, and the final HEPA filter collects any carbon fines exhausted from the charcoal adsorber.

The SGT System automatically starts and operates in response to actuation signals indicative of conditions or an accident that could require operation of the system. Following initiation, both SGT subsystem fans start. Upon verification that both subsystems are operating, one subsystem is normally shut down.

APPLICABLE SAFETY ANALYSES The design basis for the SGT System is to mitigate the consequences of a loss of coolant accident and refueling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 96 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) (Ref. 3). For all events analyzed, the SGT System is shown to be automatically initiated to reduce. via filtration and adsorption, the radioactive material released to the environment.

The SGT System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).

LCO Following a DBA, a minimum of one SGT subsystem is required to maintain the secondary containment at a negative pressure with respect to the environment and to process gaseous releases. Meeting the LCO requirements for two OPERABLE subsystems ensures operation of at least one SGT subsystem in the event of a single active failure. An OPERABLE SGT subsystem consists of a demister, heater, prefilter, HEPA filter, charcoal adsorber, a final HEPA filter, centrifugal fan, and associated ductwork, dampers, valves and controls.

BACKGROUND (continued)

(continued)

SGT System B 3.6.4.3 BASES JAFNPP B 3.6.4.3-3 Revision APLLICABILITY 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 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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 During movement of recently irradiated fuel assemblies, in the secondary containment, 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 (continued)

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CREVAS System B 3.7.3 BASES JAFNPP B 3.7.3-4 Revision 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 during movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay, the CREVAS 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 within the previous 96 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br />).

ACTIONS A.1 With one CREVAS subsystem inoperable for reasons other than an inoperable CRE boundary, the inoperable CREVAS subsystem must be restored to OPERABLE status within 7 days. With the plant in this condition, the remaining OPERABLE CREVAS subsystem is adequate to perform the CRE occupant protection function. However, the overall reliability is reduced because a failure in the OPERABLE subsystem could result in loss of the CREVAS System function. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and that the remaining subsystem can provide the required capabilities.

B.1, B.2, and B.3 If the unfiltered inleakage of potentially contaminated air past the CRE boundary and into the CRE can result in CRE occupant radiological dose greater than the calculated dose of the licensing basis analyses of DBA consequences (allowed to be up to 5 rem whole body or its equivalent to any part of the body), or inadequate protection of CRE occupants from hazardous chemicals or smoke, the CRE boundary is inoperable. Actions must be taken to restore an OPERABLE CRE boundary within 90 days.

During the period that the CRE boundary is considered inoperable, action must be initiated to implement mitigating actions to lessen the effect on CRE occupants from the potential hazards of a radiological or chemical event or a challenge from smoke. Actions must be taken within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to verify that in the event of a DBA, the mitigating actions will ensure that CRE occupant radiological exposures will not (continued)

(continued)

Control Room AC System B 3.7.4 BASES JAFNPP B 3.7.4-2 Revision Room AC System maintains a habitable environment and ensures the OPERABILITY of components in the control room. A single active component failure of a component of the Control Room AC System, assuming a loss of offsite power, does not impair the ability of the 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 coils 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 during 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 96 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br />).

(continued)

APPLICABLE SAFETY ANALYSIS (continued)

AC Sources - Shutdown B 3.8.2 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources - Shutdown BASES JAFNPP B 3.8.2-1 Revision BACKGROUND 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 SAFETY ANALYSIS The OPERABILITY of the minimum AC 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 AC electrical power is provided to mitigate events postulated during shutdown, such as 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 96 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br />).

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 analyses in MODES 4 and 5.

(continued)

AC Sources - Shutdown B 3.8.2 BASES JAFNPP B 3.8.2-4 Revision 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 are available;

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 96 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br />) 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.

A.1 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 LCO (continued)

(continued)

DC Sources - Shutdown B 3.8.5 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC Sources - Shutdown BASES JAFNPP B 3.8.5-1 Revision BACKGROUND A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC Sources Operating."

APPLICABLE SAFETY ANALYSIS The initial conditions of Design Basis Accident and 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 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 96 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br />).

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 concurrent loss of all offsite or 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 (continued)

DC Sources - Shutdown B 3.8.5 BASES JAFNPP B 3.8.5-3 Revision APPLICABILITY The DC electrical power sources required to be OPERABLE inMODES 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 are available;

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 10496 hours) 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 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 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 secondary containment). Suspension of these (continued)

(continued)

Distribution Systems - Shutdown B 3.8.8 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Distribution Systems - Shutdown BASES JAFNPP B 3.8.8-1 Revision BACKGROUND A description of the AC and 125 VDC electrical power distribution system is provided in the Bases for LCO 3.8.7, "Distribution Systems - Operating."

APPLICABLE SAFETY ANALYSIS The initial conditions of Design Basis Accident and 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 a fuel handling accident involving handling recently irradiated fuel. Due to radioactive decay, AC and 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 96 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br />).

The AC and 125 VDC electrical power distribution systems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 3).

(continued)

Distribution Systems - Shutdown B 3.8.8 BASES JAFNPP B 3.8.8-2 Revision 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).

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 are available;

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 96 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br />) 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.

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, (continued)

(continued)