RS-15-002, License Amendment Request to Revise Technical Specifications Sections 3.5.1, ECCS - Operating

License Amendment Request to Revise Technical Specifications Sections 3.5.1, ECCS - Operating

ML15012A544
Person / Time
Site:	LaSalle
Issue date:	01/12/2015
From:	Gullott D Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-15-002
Download: ML15012A544 (15)

Text

4300 Winfield Road Warrenville, IL 60555 AP"- Exelon Generation, 630 657 2000 Office RS-15-002 10 CFR 50.90 January 12, 2015 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374

Subject:

License Amendment Request to Revise Technical Specifications Sections 3.5.1, "ECCS Operating" In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit or early site permit," Exelon Generation Company, LLC (EGG) requests an amendment to Facility Operating License Nos. NPF-11 and NPF-18 for LaSalle County Station (LSCS), Units 1 and 2, respectively.

The proposed amendment would delete the Limiting Condition for Operation (LCO) Note associated with Technical Specifications (TS) Section 3.5.1, "ECCS Operating," to reflect the Residual Heat Removal (RHR) system design and ensure the RHR system operation consistent with the TS Section 3.5.1 LCO requirements.

The attached amendment request is subdivided as follows:

Attachment 1 provides a description and evaluation of the proposed change.

Attachment 2 provides the marked-up TS page with the proposed change indicated.

Attachment 3 provides the marked-up TS Bases page with the proposed change indicated. The TS Bases pages are provided for information only and do not require NRC approval.

The proposed amendment has been reviewed by the LSCS Plant Operations Review Committee and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGG Quality Assurance Program.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), EGG is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State of Illinois official.

EGC requests approval of the proposed license amendment by January 12, 2016. Once approved, the amendment will be implemented within 60 days of issuance.

January 12, 2015 U. S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Dwi Murray at (630) 657-3695.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 12th day of January 2015.

Respectfully, David M. Gullott Manager Licensing Exelon Generation Company, LLC Attachments:

1. Evaluation of Proposed Change

2. Proposed Technical Specifications Changes for LaSalle County Station, Units 1 and 2

3. Proposed Technical Specifications Bases Changes for LaSalle County Station, Units 1 and 2 cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector, LaSalle County Station NRC Project Manager, NRR LaSalle County Station Illinois Emergency Management Agency Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Change

Subject:

License Amendment Request to Revise Technical Specifications Sections 3.5.1, "ECCS Operating" 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

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

6.0 REFERENCES

Page 1 of 9

ATTACHMENT 1 Evaluation of Proposed Change 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit or early site permit," Exelon Generation Company, LLC (EGC) is requesting to amend Facility Operating License Nos. NPF-11 and NPF-18 for LaSalle County Station (LSCS), Units 1 and 2, respectively.

EGO proposes to delete the Limiting Condition for Operation (LCO) Note associated with Technical Specifications (TS) Section 3.5.1, "ECCS Operating," to reflect the Residual Heat Removal (RHR) system design and ensure the RHR system operation consistent with the TS Section 3.5.1 LCO requirements.

2.0 DETAILED DESCRIPTION The proposed change will delete the following LCO Note associated with TS 3.5.1:

NOTE: Low pressure coolant injection (LPCI) subsystems may be OPERABLE during alignment and operation for decay heat removal with reactor vessel pressure less than the residual heat removal cut-in permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable. provides the marked-up TS page with the proposed change indicated.

Attachment 3 provides the marked-up TS Bases page with the proposed change indicated for LSCS and is provided for information only.

3.0 TECHNICAL EVALUATION

RHR System Design and Operation The LSCS RHR system is designed to perform different and independent functions to support plant operation. As described in the LSCS Updated Final Safety Analysis Report (UFSAR), the safety function of the RHR system is to cool the reactor core by flooding after a loss-of-coolant accident (LOCA) (i.e., low-pressure coolant injection (LPCI) mode). The RHR system is also capable of cooling the suppression pool to safely terminate a post-LOCA containment temperature transient (i.e., containment cooling mode). In addition, the RHR system may be used during normal plant operation in the Shutdown Cooling (SDC) mode to remove residual heat from the nuclear system and cool the reactor coolant system to 125°F.

The RHR system consists of three independent closed loops, each containing a motor driven pump, powered by an ESF bus, and associated piping, valves, instrumentation and controls.

Two of the independent RHR loops contain a heat exchanger with associated service water supply system to support the heat removal functions. The RHR pumps are sized on the basis of the flow required during the LPCI mode of operation. The heat exchangers are sized on the basis of required duty for the shutdown cooling mode.

Page 2 of 9

ATTACHMENT 1 Evaluation of Proposed Change The emergency core cooling system (ECCS) systems consist of the high-pressure core spray (HPCS) system, the low-pressure core spray (LPCS) system, LPCI subsystems, and the automatic depressurization system (ADS). The LPCI mode of RHR operation supports the ECCS safety objective to limit the release of radioactive materials following a LOCA. The LPCI function is capable of delivering a large flood of water into the core to refill the reactor pressure vessel (RPV) and provide core cooling at low RPV pressures. The three RHR pumps automatically start in LPCI mode upon receipt of an ECCS initiation signal. LPCI is a low-head, high-flow function that delivers flow to the RPV when the differential pressure between the RPV and drywell is less than 225 psid (rated flow is injected at 20 psid). LPCI is designed to reflood the RPV to at least two-thirds core height and to maintain this level. In the LPCI mode of operation, each RHR pump takes suction from the suppression pool through an independent suction line and discharge to the reactor core through separate RPV piping penetrations.

The SDC mode of the RHR system is operated during normal unit cooldown and shutdown to remove decay heat. Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant. This decay heat removal is required for performing refueling or maintenance operations, or for keeping the reactor in the Mode 3 (i.e., Hot Shutdown) condition. The initial phase of nuclear system cooldown is accomplished by dumping steam from the RPV to the main condenser. When nuclear system temperature has decreased to where the steam supply pressure is not sufficient to maintain the turbine shaft gland seals, vacuum in the main condenser cannot be maintained and the RHR system is placed in the SDC mode of operation. The SDC subsystem is able to remove decay heat to complete cooldown to 125°F in less than 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> after the control rods have been inserted and can maintain the nuclear system at or below 125°F. In the SDC mode of operation, the 'A' or '13' RHR pump takes suction from the `A' Reactor Recirculation (RR) loop; and directs the flow through the RHR heat exchanger prior to returning the water back to the RPV through RR pump discharge piping.

LPCI and SDC TS Requirements LSCS TS 3.5.1, "ECCS-Operating," requires each ECCS injection system be OPERABLE in Modes 1, 2, and 3. With regard to the LPCI mode, this means that all three LPCI subsystems are required for the LCO to be met. If one LPCI subsystem is inoperable, TS 3.5.1, Condition A requires the inoperable subsystem to be returned to OPERABLE status within seven days.

The TS 3.5.1 LCO is modified by a note that allows a LPCI subsystem to be considered OPERABLE for the LPCI function when the subsystem is being aligned or is operating in the SDC mode, and the unit is in Mode 3 below the RHR cut in permissive pressure. Utilization of this note requires that the RHR system be capable of manual realignment to the LPCI mode and not be otherwise inoperable. The note was added to TS LCO 3.5.1 during LSCS Improved Technical Specifications (ITS) conversion as part of TS Amendment 147 and 133 for Units 1 and 2, respectively, which was approved on March 30, 2001 (Reference 1). The allowance provided by the note was considered acceptable because the return to operability entails only the repositioning of valves, either remote or locally, and the energy requiring dissipation in Mode 3, below the RHR cut-in permissive pressure is considerably less than that at 100% power with normal operating temperature and pressure.

Page 3 of 9

ATTACHMENT 1 Evaluation of Proposed Change LSCS TS 3.4.9, "Residual Heat Removal (RHR) Shutdown Cooling System Hot Shutdown,"

LCO requires that two RHR SDC subsystems to be OPERABLE; and when no Reactor Recirculation pump is in operation, one SDC subsystem must be in operation during Mode 3. An OPERABLE RHR SDC subsystem consists of one RHR pump, one heat exchanger, and the associated piping and valves. Each SDC subsystem is considered OPERABLE if it can be manually aligned (remote or local) to the SDC mode for removal of decay heat. In Mode 3, one RHR SDC subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. If one or both SDC subsystems are inoperable, TS 3.4.9, Condition A, requires immediate initiation of actions to restore one SDC subsystem to OPERABLE status.

TS 3.4.9 is only applicable in Mode 3, with RPV pressure less than the RHR cut-in permissive pressure. Mode 3 means that the reactor mode switch is in the 'Shutdown' position and the average reactor coolant temperature is greater than 200°F. In this mode, all the RPV head closure bolts are fully tensioned and therefore, RPV pressure is typically above atmospheric pressure.

LPCI and SDC Operation During an NRC inspection conducted in 2012 in accordance with Temporary Instruction (TI) 2515/177, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems (NRC Generic Letter 2008-01)," NRC inspectors identified a concern because the operability of LPCI was not ensured in Mode 3 while an RHR subsystem was operating in SDC mode as required by TS. The Inspection Report (Reference 2) stated that TS LCO 3.5.1 required, in part, each ECCS subsystem be operable during Mode 3 and that this LCO included the Note discussed above. Contrary to the requirements of this LCO 3.5.1 Note, LSCS had declared the LPCI subsystem inoperable when it was configured for the SDC mode of operation.

LSCS has been operating in this manner (i.e., declaring a LPCI subsystem inoperable when configured for SDC mode of operation) since 1995 when procedural restrictions preventing the realignment of an RHR subsystem from SDC mode to LPCI mode were implemented. The procedural restrictions resulted from LSCS specific operating experience which identified that the LPCI suction valve from the Suppression Pool was susceptible to thermal binding. This condition exists when the differential temperature across the valve is greater than or equal to 60°F, as it may be when the valve is closed to support the SDC mode of operation. This temperature differential results from the high temperature SDC water on one side of the valve and the cool suppression pool water on the other.

Subsequent to these procedural restrictions, an LSCS review of NRC Information Notice (IN) 2010-11 (Reference 3) and operating experience at Prairie Island Licensee Event Report (LER) 1-09-04 (Reference 4), determined that during operation in Mode 3, the potential exists for the water in the RHR pump suction piping aligned for SDC to flash/boil when realigned to the LPCI mode. This phenomenon is due to the physical arrangement (i.e., common interface) of the SDC and LPCI suction lines for the RHR pumps. The realignment from SDC mode to LPCI mode transfers the suction source for the RHR pump; thereby exposing the high temperature SDC water to the low pressure LPCI suction piping from the Suppression Pool. The resultant flashing/boiling of the high pressure, high temperature water when introduced to the low Page 4 of 9

ATTACHMENT 1 Evaluation of Proposed Change pressure piping could result in voiding in the suction piping, RHR pump cavitation, water hammer and associated RHR system damage. This vulnerability is greatest during the early stages of Mode 3 operation when the SDC water temperature is highest.

The flashing/boiling in the RHR suction piping and the Suppression Pool suction valve thermal binding are the result of the RHR system design that supports several different operating modes using common equipment. This design feature, and the associated temperature phenomenon, prevents timely realignment of the RHR subsystem from SDC mode to LPCI mode. Therefore, the TS 3.5.1 Note that allows an RHR subsystem to remain OPERABLE for LPCI mode when being aligned or operated in SDC mode is inappropriate and should be removed from the LSCS TS. LSCS will continue to declare the respective LPCI subsystem of ECCS inoperable for the subsystem operating in SDC mode and enter the appropriate Condition(s) of TS 3.5.1 in Mode

3. This operation is consistent with LSCS practice in declaring the respective containment cooling modes of RHR inoperable for the RHR subsystem operating in SDC mode and enter the appropriate Condition(s) of TS Sections 3.6.2.3, "RHR Suppression Pool Cooling," and 3.6.2.4, "RHR Suppression Pool Spray," during Mode 3.

Justification for Note Removal Plant operation described above (i.e., RHR subsystem in SDC mode and not available to support LPCI mode) is consistent with the original LSCS RHR system design and approval.

NUREG-0519, "Safety Evaluation Report related to the operation of LaSalle County Station Units 1 and 2," (Reference 5) reviewed and approved the LSCS RHR system design, noting that the LPCI and SDC modes of operation are mutually exclusive for a selected subsystem. The low energy conditions in Mode 3, with one RHR subsystem aligned in SDC and not capable of performing its LPCI function, were evaluated during initial LSCS licensing. The evaluations confirmed significantly greater margin to peak cladding temperature limits for Mode 3 LOCA events than those from full power. This result is due to the stored sensible heat in the fuel and reactor pressure vessel, and the enthalpy of the reactor coolant, being significantly reduced when the RHR cut-in permissive has been reached in Mode 3. Additionally, the reactor core decay heat is reduced to a small fraction of the level immediately after sub-criticality has been achieved.

The NRC approved Standard Technical Specifications (STS - NUREG-1434, Revision 4) also recognized this boiling water reactor design configuration and the mutual exclusivity of the LPCI and SDC functions. The STS bases explained the allowance provided by the TS 3.5.1 Note as "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. 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." However, industry and site specific operating experience makes the application of the TS 3.5.1 note inappropriate at LSCS.

Therefore, the removal of the TS 3.5.1 note, and operation with one RHR subsystem inoperable for LPCI mode while being aligned or operated in SDC in accordance with TS 3.4.9, is justified.

Page 5 of 9

ATTACHMENT 1 Evaluation of Proposed Change

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The following NRC requirements and guidance documents are applicable to the review of the proposed changes.

10 CFR 50, Appendix A, General Design Criterion (GDC) 34, Residual heat removal," requires that a system to remove residual heat be provided with a safety function to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

10 CFR 50, Appendix A, GDC 35, "Emergency core cooling," requires that a system to provide abundant emergency core cooling be provided with a safety function to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

10 CFR 10, Appendix A, GDC 37, "Testing of emergency core cooling system," requires that the emergency core cooling system design provide the capability for periodic pressure and functional testing. This testing shall assure (1) structural and leaktight integrity of components, (2) operability and performance of active components, (3) operability of the whole system under conditions as close to design as possible.

10 CFR 50.36, "Technical specifications," details the content and information that must be included in a station's Technical Specifications (TS). In accordance with 10 CFR 50.36, IS are required to include (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. As described in 10 CFR 50.36(c)(2), "Limiting conditions for operation,"

are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO is not met, the licensee shall shut down the reactor or follow any other actions permitted by IS.

10 CFR 50.46(a)(1)(i) requires that each boiling or pressurized light-water nuclear power reactor be provided with an ECCS designed with a calculated cooling performance in accordance with an acceptable evaluation model following a postulated LOCA.

The proposed change does not involve any physical changes to the structures, systems, or components at LSCS. The proposed change will reflect current plant configuration of the RHR system design and assure safe operation by continuing to meet applicable regulations and requirements.

4.2 Precedents The NRC has approved a similar license amendment request to remove this note from the IS for emergency core cooling system as follows:

Page 6 of 9

ATTACHMENT 1 Evaluation of Proposed Change Letter from R. B. Ennis (NRC) to M. J. Pacilio (EGO), "Peach Bottom Atomic Power Station, Units 2 and 3 Issuance of Amendments Re: Delete Non-Conservative Note from Limiting Condition for Operation for Operation 3.5.1 (TAO Nos. MF3184 and MF3185)," dated July 28, 2014 (ADAMS Accession Number ML14163A589).

4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGO) requests and amendment to Facility Operating License Nos. NPF-11 and NPF-18 for LaSalle County Station (LSCS), Units 1 and 2.

The proposed amendment would delete the Limiting Condition for Operation (LCO) Note associated with Technical Specifications (TS) Section 3.5.1, "ECCS Operating," to reflect current plant configuration and ensure the Residual Heat Removal (RHR) system operation consistent with the TS Section 3.5.1 LCO requirements.

EGO has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92(c),

"Issuance of amendment," as discussed below:

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

Response: No.

No physical changes to the facility will occur as a result of this proposed amendment. The proposed change will not alter the physical design. Current TS note could make LSCS susceptible to potential water hammer in the RHR system if in the SDC Mode of RHR in Mode 3 when swapping from the SDC to LPCI mode of RHR. The proposed LAR will eliminate the risk for cavitation of the pump and voiding in the suction piping, thereby avoiding potential to damage the RHR system, including water hammer.

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

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

Response: No.

The proposed change does not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Accordingly, the change does not introduce any new accident initiators, nor does it reduce or adversely affect the capabilities of any plant structure, system, or component to perform their safety function. Deletion of the TS Note is appropriate because current TSs could put the plant at risk for potential cavitation of the pump and voiding in the suction piping, resulting in potential to damage the RHR system, including water hammer.

Page 7 of 9

ATTACHMENT 1 Evaluation of Proposed Change 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 a margin of safety?

Response: No.

The proposed change conforms to NRC regulatory guidance regarding the content of plant Technical Specifications. The proposed change does not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant.

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

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

4.4 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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

EGC has evaluated the proposed amendment for environmental considerations. The review has resulted in the determination that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any 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(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.

6.0 REFERENCES

1. Letter from S. N. Bailey (U. S. Nuclear Regulatory Commission) to 0. D. Kingsley (Exelon Generation Company, LLC), "Issuance of Amendments (TAC Nos. MA8388 and MA8390)," dated March 30, 2001.

Page 8 of 9

ATTACHMENT 1 Evaluation of Proposed Change

2. Letter from M. Kunowski (U. S. Nuclear Regulatory Commission) to M. J. Pacilio (Exelon Generation Company, LLC), "LaSalle County Station, Units 1 and 2 NRC Integrated Inspection Report 05000373/2012004; 05000374/2012004," dated October 30, 2012.

3. NRC Information Notice 2010-11, "Potential for Steam Voiding Causing Residual Heat Removal System Inoperability," dated June 16, 2010.

4. Prairie Island Nuclear Generating Plant LER 1-09-04, "Residual Heat Removal System Inoperability While in Mode 4 Due to Potential Steam Voiding," dated June 5, 2009.

5. LaSalle County Station, Units 1 and 2 Safety Evaluation Report (SER), NUREG-0519, dated March 1981.

Page 9 of 9

ATTACHMENT 2 Proposed Technical Specifications Changes for LaSalle County Station, Units 1 and 2 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 AFFECTED TECHNICAL SPECIFICATIONS PAGE 3.5.1 1

ECCS Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 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 h at removal with reactor vessel prcssurc less than thc residual heat removal cut in permissive prcssurc in MODE 3, if capable of being manually re-aligned and not otherwise inoperable.

APPLICABILITY: MODE 1, MODES 2 and 3, except 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 HPCS.

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 to OPERABLE status.

(continued)

LaSalle 1 and 2 3.5.1-1 Amendment No. 171/157

ATTACHMENT 3 Proposed Technical Specifications Bases Changes for LaSalle County Station, Units 1 and 2 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 AFFECTED TECHNICAL SPECIFICATIONS BASES PAGE (NOTE: TS Bases pages are provided for information only.)

B 3.5.1 5

ECCS Operating B 3.5.1 BASES APPLICABLE the most severe failure. The remaining OPERABLE ECCS SAFETY ANALYSES subsystems, which include one spray subsystem, provide the (continued) capability to adequately cool the core, under near-term and long-term conditions, and prevent excessive fuel damage.

For all LOCA analyses, only six ADS valves are assumed to function. An additional analysis has been performed which assumes five ADS valves function, however in this analysis all low pressure and high pressure ECCS subsystems are also assumed to be available.

The ECCS satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Each ECCS injection/spray subsystem and six ADS valves are required to be OPERABLE. The ECCS injection/spray subsystems are defined as the three LPCI subsystems, the LPCS System, and the HPCS System. The low pressure ECCS injection/spray subsystems are defined as the LPCS System and the three LPCI subsystems.

With less than the required number of ECCS subsystems OPERABLE during a limiting design basis LOCA concurrent with the worst case single failure, the limits specified in 10 CFR 50.46 (Ref. 10) could potentially be exceeded. All ECCS subsystems must therefore be OPERABLE to satisfy the single failur riterion required by 10 CFR 50. . 10).

As notcd, LPCI subsystem& may bc considered OPERABLE during alignment and operation for decay heat remov when below thc actual RHR cut in permissive pressure ODE 3, if capable of bcing manuaTly realigned (r - *tc or local) to thc LPCI modc d ot oth:,r2 . Alignment an-et.

operation for accoywat rem3va ncludes: a) W-19-&19-system 's realigned to o-F-4ro thc RHR shutdo.,;- eoo'ing mode b) bon thc system . 51n thc RHR shutdown cooing mode, w-het-19-e-e or not the R. pump is operating. This allowance is necessary since RHR System may be.e.cLirec to oer4-t e i rl_

the shutdown ..oling mode to remove decay heat an s 7.'i,+e heat from e reactor. At these low pressures and decay heat 1-fels, a reduced complement of ECCS subsystems should pri,,ide the required core cooling, thereby allowing peration of RHR shutdown cooling when necessary.

(continued)

!since transferring from the shutdown cooling mode to the LPCI mode

,could result in pump cavitation and voiding in the suction piping, resulting in the potential to damage the RHR System, including water

'hammer.

LaSalle 1 and 2 B 3.5.1-5 Revision 4§