L-MT-18-009, License Amendment Request: Revise Limiting Condition for Operation (LCO) of Specification 3.5.1, Emergency Core Cooling System - Operating, to Remove the LCO Note

License Amendment Request: Revise Limiting Condition for Operation (LCO) of Specification 3.5.1, Emergency Core Cooling System - Operating, to Remove the LCO Note

ML18317A172
Person / Time
Site:	Monticello
Issue date:	11/12/2018
From:	Church C Northern States Power Company, Minnesota
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-18-009
Download: ML18317A172 (18)

Text

2807 West County Road 75 Monticello, MN 55362 800.895.4999 xcelenergy.com November 12, 2018 L-MT-18-009 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 License Amendment Request: Revise Limiting Condition for Operation (LCO) of Specification 3.5.1, Emergency Core Cooling System - Operating, to Remove the LCO Note In accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, the Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requests an amendment to the Technical Specifications (TS) for the Monticello Nuclear Generating Plant (MNGP). The proposed TS change revises the Limiting Condition for Operation (LCO) of Specification 3.5.1, "[Emergency Core Cooling System] ECCS - Operating", to remove the LCO note.

Enclosed is a description and assessment of the proposed TS change. The enclosure also provides the no significant hazards consideration evaluation in accordance with 10 CFR 50.92, Issuance of amendment, and the environmental assessment. These provide the bases for the conclusion that the license amendment request involves no significant hazards consideration and meets the eligibility criterion for a categorical exclusion as set forth in 10 CFR 51.22, Criteria for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review, specifically paragraph (c)(9). to the enclosure provides the existing TS page marked up to show the proposed change. Attachment 2 to the enclosure provides the revised (clean) TS page. Attachment 3 to the enclosure provides the TS Bases page marked up to show the associated TS bases changes and is provided for information only.

NSPM requests NRC approval of the proposed license amendment request by December 12, 2019. Once approved, the amendment will be implemented within 90 days.

Document Control Desk Page 2 In accordance with 10 CFR 50.91, "Notice for public comment; State consultation",

paragraph (b), NSPM is notifying the State of Minnesota by providing a copy of this application, with this enclosure and attachments, to the designated State Official.

If additional information is needed, please contact Mr. Richard Loeffler at (612) 342-8981.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

I declare, under penalty of perjury, that the foregoing is true and correct. Executed on November 11._, 2018.

-4.~

hristopher R. Church Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc: Administrator, Region Ill, US NRC Project Manager, Monticello, US NRC Resident Inspector, Monticello, US NRC State of Minnesota

ENCLOSURE MONTICELLO NUCLEAR GENERATING PLANT EVALUATION OF THE PROPOSED CHANGE REVISE LIMITING CONDITION FOR OPERATION (LCO) OF SPECIFICATION 3.5.1, EMERGENCY CORE COOLING SYSTEM - OPERATING, TO REMOVE THE LCO NOTE 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Description and Reason for the Proposed Change 2.3 Facility Description 2.4 Impact on Submittals Under NRC Review

4.0 REGULATORY ANALYSIS

4.1 Applicable Regulatory Requirements 4.2 Precedents 4.3 No Significant Hazards Consideration Analysis 4.4 Conclusions 5.0 ENVIRONMENTAL EVALUATION

6.0 REFERENCES

ATTACHMENTS:

1. Technical Specification Pages (Mark-up)

2. Technical Specification Pages (Retyped)

3. Technical Specification Bases Pages (Mark-up - for information only)

Page 1 of 10

L-MT-18-009 NSPM Enclosure REVISE LIMITING CONDITION FOR OPERATION (LCO) OF SPECIFICATION 3.5.1, EMERGENCY CORE COOLING SYSTEM - OPERATING, TO REMOVE THE LCO NOTE 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, the Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requests to revise the Technical Specifications (TS) for the Monticello Nuclear Generating Plant (MNGP). The proposed TS change revises the Limiting Condition for Operation (LCO) of Specification 3.5.1, "[Emergency Core Cooling System] ECCS - Operating", by removing a note to reflect the Residual Heat Removal (RHR)

System design and to ensure system operation is consistent with LCO requirements.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation As described in the MNGP Updated Safety Analysis Report (USAR), Section 6.2.3, Residual Heat Removal System (RHR), and Section 10.2.4, Reactor Shutdown Cooling System, the RHR System is designed to perform several different and independent safety and operational objectives to support plant operation. The Low Pressure Coolant Injection (LPCI) and Shutdown Cooling (SDC) modes are pertinent to this LAR and are discussed in more detail below. The LPCI mode restores and maintains coolant inventory in the reactor vessel so that the core is adequately cooled after a Loss-of-Coolant Accident (LOCA). The SDC mode removes decay and residual heat from the nuclear system to allow performance of refueling and primary system servicing.

RHR System Design The RHR System consists of two subsystems. Each RHR subsystem, consisting of a heat exchanger, two RHR pumps in parallel and associated piping, is located in a different corner room of the lowest elevation of the Reactor Building. The discharge of the two RHR subsystems is cross connected by a single header, making it possible to supply either RHR subsystem from the pumps in the other subsystem. In the LPCI mode, each LPCI subsystem consists of a common suction line from the suppression pool, parallel flowpaths through the two RHR pumps, and a common injection line to the reactor pressure vessel (RPV). In the SDC mode, a subsystem consists of a common suction line from the No. 11 reactor recirculation loop, parallel flowpaths through the RHR pump(s), through the RHR heat exchanger to discharge heat to the RHR Service Water System, and then return to the RPV through either one of the recirculation loops.

Page 2 of 10

L-MT-18-009 NSPM Enclosure LPCI Mode The ECCS network consists of the High Pressure Coolant Injection System, the Core Spray System, the LPCI mode of the RHR System, and the Automatic Depressurization System.

The LPCI subsystems are designed to provide core cooling at low RPV pressure. LPCI (a low head, high flow system) delivers rated flow to the RPV when the differential pressure between the RPV and primary containment is approximately 20 psi or less. During operation in the LPCI mode, the RHR pumps take suction from the suppression pool and discharge to the RPV through the recirculation loop selected by the LPCI loop selection logic. The LPCI mode is designed to reflood and maintain the RPV water level to at least two-thirds core height.

SDC Mode The SDC mode is designed to remove the decay and sensible heat during a normal reactor shutdown and cooldown. This operational mode is manually initiated at low reactor pressure and is procedurally controlled to preclude water hammer transients. The initial phase of the nuclear system cooldown is accomplished by dumping steam from the RPV to the main condenser with the condenser acting as the heat sink. When the nuclear system temperature has decreased to a point where the steam supply pressure is insufficient 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.

2.2 Description and Reason for the Proposed Change The proposed TS change removes the following note in the LCO of Specification 3.5.1 to reflect the RHR System design and ensure system operation is consistent with LCO requirements.

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) shutdown cooling supply isolation interlock in MODE 3, if capable of being manually realigned and not otherwise inoperable. to the enclosure provides the existing TS page marked up to show the proposed changes. Attachment 2 to the enclosure provides the revised (clean) TS page. Attachment 3 to the enclosure provides the TS Bases page marked up to show the associated TS Bases changes and is provided for information only. TS Bases changes are issued in accordance with MNGP Specification 5.5.9, Technical Specification (TS) Bases Control Program, following approval of the associated license amendment request (LAR).

Page 3 of 10

L-MT-18-009 NSPM Enclosure 2.3 Facility Description MNGP is a single unit plant located on the south bank of the Mississippi River within the city limits of Monticello, Minnesota. The MNGP is a single cycle, forced circulation, low power density boiling water reactor, designed and supplied by the General Electric Corporation. The MNGP application for a Construction Permit and Operating License was submitted to the Atomic Energy Commission (AEC) on August 1, 1966. Amendment No. 1 to Provisional Operating License No. DPR-22 was issued on January 13, 1971, granting full power operation.

MNGP began full power commercial operation on June 30, 1971.

The MNGP was designed and constructed to comply with NSPMs understanding of the intent of the AEC General Design Criteria (GDC) for Nuclear Power Plant Construction Permits, as published on July 11, 1967. MNGP was not licensed to NUREG-0800, Standard Review Plan.

2.4 Impact on Submittals Under NRC Review There is no impact on any submittals currently under NRC review since the proposed TS changes do not involve any specifications that are proposed to be modified in the other submittals.

3.0 TECHNICAL EVALUATION

Specification 3.5.1, "ECCS-Operating", requires each ECCS injection / spray system to be operable in Modes 1, 2, and 3. This requires both LPCI subsystems to be operable to meet the LCO. Specification 3.4.7, "Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown", LCO requires two RHR SDC subsystems to be operable, and with no recirculation pump in operation, at least one RHR SDC subsystem to be in operation. Specification 3.4.7 is applicable in Mode 3 (hot shutdown) when the reactor steam dome pressure is less than the RHR shutdown cooling supply isolation interlock pressure. This interlock is provided to prevent the overpressurization of the RHR suction line while in SDC by inadvertent operation of valves in the suction line from the recirculation loop. Operation in SDC is not permitted when the reactor steam dome pressure is greater than this interlock setting of approximately 80 psig. In hot shutdown when reactor steam dome pressure is below the RHR shutdown cooling supply isolation interlock pressure, SDC may be placed in service.

The LCO of Specification 3.5.1 contains a note stating that in hot shutdown the LPCI subsystem may be considered operable during alignment and operation for decay heat removal when the reactor steam dome pressure is less than the RHR shutdown cooling supply isolation interlock, if it is capable of being manually realigned to the LPCI mode and is not otherwise inoperable. However, industry operating experience has indicated that application of this LCO note could result in some cases in water flashing to steam in the RHR System piping, water hammer, pressure locking, and/or thermal binding of valves. This is similar to the concerns reflected in NRC Information Notice 2010-11, Potential for Steam Voiding Causing Residual Heat Removal System Inoperability, dated June 16, 2010. This phenomenon is Page 4 of 10

L-MT-18-009 NSPM Enclosure applicable to some boiling water reactor plants due to the physical arrangement (that is the common interface) of the SDC and LPCI suction lines for the RHR pumps. Realignment from SDC to the LPCI mode transfers the suction source for the RHR pump. The resultant flashing or boiling of the high pressure / high temperature SDC water when introduced to the low pressure LPCI piping could result in voiding in the suction piping, RHR pump cavitation, water hammer and potentially RHR System damage. This possibility is greatest during the early stages of hot shutdown operation when the SDC water temperature is the highest.

NSPM does not have an analysis to demonstrate that realignment of an RHR subsystem from SDC to the LPCI mode does not result in thermal-hydraulic transients which could potentially challenge the system under certain scenarios during realignment to the LPCI for injection.

Therefore, continued retention of this note is inappropriate and this note should be removed from the MNGP TS.

4.0 REGULATORY ANALYSIS

4.1 Applicable Regulatory Requirements 10 CFR 50.36, Technical specifications", details the content and information that must be included in a facilities TS. In accordance with the requirements of 10 CFR 50.36, the TS are required to include: (1) safety limits and limiting safety system 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 TS.

10 CFR 50.46(a)(1)(i), Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, 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 applicable 10 CFR 50, Appendix A, GDC criterion is presented first below followed by the corresponding criteria from the 70 draft AEC GDCs provided for comparison. The applicable Principal Design Criteria (PDC) from MNGP USAR Subsection 1.2.3, Reactor Core Cooling are then provided for comparison.

10 CFR 50, Appendix A, GDC Criterion 34, Residual heat removal, states in part:

A system to remove residual heat shall be provided. The system safety function shall be 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.

Page 5 of 10

L-MT-18-009 NSPM Enclosure The corresponding 1967 AEC proposed GDC Criterion 6, Reactor Core Design (Category A),

states:

The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of off-site power.

10 CFR 50, Appendix A, GDC Criterion 35, Emergency core cooling, states:

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

The corresponding 1967 AEC proposed GDC Criterion 44, Emergency core cooling capability (Category A), states:

At least two emergency core cooling systems, preferably of different design principles, each with a capability for accomplishing abundant emergency core cooling, shall be provided. Each emergency core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal-water reaction to negligible amounts of all sizes of breaks in the reactor coolant pressure boundary, including the double-ended rupture of the largest pipe. The performance of each emergency core cooling system shall be evaluated conservatively in each area of uncertainty. The systems shall not share active components and shall not share other features or components unless it can be demonstrated that (a) the capability of the shared feature or components to perform its required function can be readily ascertained during reactor operation, (b) failure of the shared feature or component does not initiate a loss-of-coolant accident, and (c) capability of the shared feature or component to perform its required function is not impaired by the effects of a loss-of-coolant accident and is not lost during the entire period this function is required following the accident.

Page 6 of 10

L-MT-18-009 NSPM Enclosure Meeting the intent of the 10 CFR 50, Appendix A GDCs is supported by the design of the plant to the General Electric PDCs stated below from Subsection 1.2.3, Reactor Core Cooling, of the MNGP USAR:

a. Heat removal systems are provided to remove heat generated in the reactor core for the full range of normal operational conditions from plant shutdown to maximum thermal output. The capacity of such systems is adequate to prevent fuel clad damage.

b. Heat removal systems are provided to remove decay heat generated in the reactor core under circumstances wherein the normal operational heat removal systems become inoperative. The capacity of such systems is adequate to prevent fuel clad damage.

c. Redundant heat removal systems are provided to preserve reactor core heat transfer geometry following various postulated design basis loss-of-coolant accidents.

The proposed TS change reflects the configuration of the RHR System design and does not involve any physical changes to the systems, structures and components at the MNGP. The 10 CFR 50, Appendix A GDCs, the 1967 AEC proposed draft 70 GDCs, and the MNGP PDCs, while worded somewhat differently, are equivalent in that systems to afford abundant emergency core cooling and to remove residual heat are provided. The safety functions to transfer heat from the reactor core following a LOCA and transfer fission product decay and residual heat from the reactor core continue to meet the applicable regulations and requirements.

4.2 Precedents Multiple amendments to remove this LCO 3.5.1 note (or an equivalent note contained within a surveillance requirement) have been approved by the NRC. This LAR is consistent with those amendments. The most recent amendments noted are for the Exelon Generation Company Nine Mile Point Nuclear Station, Unit 2 and for the Energy Northwest Columbia Generating Station, References 1 and 2, respectively.

4.3 No Significant Hazards Consideration Analysis In accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, the Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), requests an amendment to the Technical Specifications (TS) for the Monticello Nuclear Generating Plant (MNGP). The proposed TS change revises the Limiting Condition for Operation (LCO) of Specification 3.5.1, "[Emergency Core Cooling System] ECCS - Operating", by removing a note to more appropriately reflect the Residual Heat Removal (RHR) System design and to ensure RHR System operation is consistent with the LCO requirements.

Page 7 of 10

L-MT-18-009 NSPM Enclosure NSPM has evaluated the proposed change against the criteria of 10 CFR 50.92, Issuance of amendment, to determine if the proposed change results in any significant hazards. The following is the evaluation of each of the 10 CFR 50.92(c) criteria:

1. Does the proposed change 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 TS change.

The proposed change will not alter the physical design of the facility. The current LCO note could make the MNGP susceptible to potential water hammer in the RHR System if in the Shutdown Cooling mode of RHR in Mode 3 when swapping from SDC to the Low Pressure Coolant Injection (LPCI) mode of RHR. The proposed license amendment request will eliminate the risk for cavitation of the RHR pumps and voiding in the suction piping, thereby avoiding the 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 change 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 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. Removal of the LCO note is appropriate because this current TS allowance could put the plant at risk for potential cavitation of the RHR pumps and voiding in the suction piping, resulting in the potential occurrence of water hammer and damage to the RHR System.

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

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

Response: No The proposed change conforms to NRC regulatory guidance regarding the content of the TSs. The proposed change does not affect the capabilities of any plant structure, system, or component to perform its associated safety function. Since the safety analysis assumptions are unaffected the associated safety margins are also not Page 8 of 10

L-MT-18-009 NSPM Enclosure impacted. Removal of the LCO note is appropriate because the current TS requirement could put the plant at risk of damage to the RHR System and impact the LPCI function.

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

Based on the above, NSPM 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 Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL EVALUATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, Standards for Protection Against Radiation, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, Criteria for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review, specifically paragraph (c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

Page 9 of 10

L-MT-18-009 NSPM Enclosure

6.0 REFERENCES

1. Letter from the NRC to Exelon Generation Company, Nine Mile Point Nuclear Station, Unit 2 - Issuance of Amendment No. 170 Re: Removal of Note Associated with Surveillance requirement 3.5.1.2 (CAC No. MG0148; EPID L-2017-LLA-0294), dated June 8, 2018 (ADAMS Accession Number ML18131A291)

2. Letter from the NRC to Energy Northwest, Columbia Generating Station - Issuance of Amendment Re: Change to Technical Specification 3.5.1, ECCS - Operating (CAC No. MG0015: EPID L-2017-LLA-0277), dated May 2, 2018 (ADAMS Accession Number ML18100A199)

Page 10 of 10

ENCLOSURE ATTACHMENT 1 MONTICELLO NUCLEAR GENERATING PLANT REVISE LIMITING CONDITION FOR OPERATION (LCO) OF SPECIFICATION 3.5.1, EMERGENCY CORE COOLING SYSTEM - OPERATING, TO REMOVE THE LCO NOTE TECHNICAL SPECIFICATION PAGE (MARK-UP) 1 page follows

ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), 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 three 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) shutdown cooling supply isolation interlock 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 LPCI pump A.1 Restore LPCI pump to 30 days inoperable. OPERABLE status.

B. One LPCI subsystem B.1 Restore low pressure 7 days inoperable for reasons ECCS injection/spray other than Condition A. subsystem to OPERABLE status.

OR U

One Core Spray subsystem inoperable.

Monticello 3.5.1-1 Amendment No. 146, 198XXX

ENCLOSURE ATTACHMENT 2 MONTICELLO NUCLEAR GENERATING PLANT REVISE LIMITING CONDITION FOR OPERATION (LCO) OF SPECIFICATION 3.5.1, EMERGENCY CORE COOLING SYSTEM - OPERATING, TO REMOVE THE LCO NOTE TECHNICAL SPECIFICATION PAGE (RETYPED) 1 page follows

ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS), 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 three safety/relief valves shall be OPERABLE.

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 LPCI pump A.1 Restore LPCI pump to 30 days inoperable. OPERABLE status.

B. One LPCI subsystem B.1 Restore low pressure 7 days inoperable for reasons ECCS injection/spray other than Condition A. subsystem to OPERABLE status.

OR U

One Core Spray subsystem inoperable.

Monticello 3.5.1-1 Amendment No. XXX

ENCLOSURE ATTACHMENT 3 MONTICELLO NUCLEAR GENERATING PLANT REVISE LIMITING CONDITION FOR OPERATION (LCO) OF SPECIFICATION 3.5.1, EMERGENCY CORE COOLING SYSTEM - OPERATING, TO REMOVE THE LCO NOTE TECHNICAL SPECIFICATION BASES PAGE (MARK-UP - FOR INFORMATION ONLY) 1 page follows

(&&62SHUDWLQJ





A is not 

%$6(6



/&2 FRQWLQXHG  $VQRWHG/3&,VXEV\VWHPVPD\EHFRQVLGHUHG23(5$%/(GXULQJ

DOLJQPHQWDQGRSHUDWLRQIRUGHFD\KHDWUHPRYDOZKHQEHORZWKHDFWXDO

since transferring from 5+5VKXWGRZQFRROLQJVXSSO\LVRODWLRQLQWHUORFNLQ02'(LIFDSDEOHRI

the shutdown cooling EHLQJPDQXDOO\UHDOLJQHG UHPRWHRUORFDO WRWKH/3&,PRGHDQGQRW

mode to the LPCI mode RWKHUZLVHLQRSHUDEOH$OLJQPHQWDQGRSHUDWLRQIRUGHFD\KHDWUHPRYDO

could result in pump LQFOXGHVZKHQWKHUHTXLUHG5+5SXPSLVQRWRSHUDWLQJRUZKHQWKH

cavitation and voiding in V\VWHPLVUHDOLJQHGIURPRUWRWKH5+5VKXWGRZQFRROLQJPRGH7KLV

the suction piping, DOORZDQFHLVQHFHVVDU\VLQFHWKH5+56\VWHPPD\EHUHTXLUHGWR

resulting in the potential RSHUDWHLQWKHVKXWGRZQFRROLQJPRGHWRUHPRYHGHFD\KHDWDQGVHQVLEOH

KHDWIURPWKHUHDFWRU$WWKHVHORZSUHVVXUHVDQGGHFD\KHDWOHYHOVD

to damage the RHR UHGXFHGFRPSOHPHQWRI(&&6VXEV\VWHPVVKRXOGSURYLGHWKHUHTXLUHG

System, including water FRUHFRROLQJWKHUHE\DOORZLQJRSHUDWLRQRI5+5VKXWGRZQFRROLQJZKHQ

hammer. QHFHVVDU\ is





$33/,&$%,/,7< $OO(&&6VXEV\VWHPVDUHUHTXLUHGWREH23(5$%/(GXULQJ02'(6

DQGZKHQWKHUHLVFRQVLGHUDEOHHQHUJ\LQWKHUHDFWRUFRUHDQGFRUH

FRROLQJZRXOGEHUHTXLUHGWRSUHYHQWIXHOGDPDJHLQWKHHYHQWRIDEUHDN

LQWKHSULPDU\V\VWHPSLSLQJ,Q02'(6DQGZKHQUHDFWRUVWHDP

GRPHSUHVVXUHLVSVLJ$'6DQG+3&,DUHQRWUHTXLUHGWREH

23(5$%/(EHFDXVHWKHORZSUHVVXUH(&&6VXEV\VWHPVFDQSURYLGH

VXIILFLHQWIORZEHORZWKLVSUHVVXUH(&&6UHTXLUHPHQWVIRU02'(6

DQGDUHVSHFLILHGLQ/&2(&&66KXWGRZQ





$&7,216 $1RWHSURKLELWVWKHDSSOLFDWLRQRI/&2EWRDQLQRSHUDEOH+3&,

VXEV\VWHP7KHUHLVDQLQFUHDVHGULVNDVVRFLDWHGZLWKHQWHULQJD02'(

RURWKHUVSHFLILHGFRQGLWLRQLQWKH$SSOLFDELOLW\ZLWKDQLQRSHUDEOH+3&,

VXEV\VWHPDQGWKHSURYLVLRQVRI/&2EZKLFKDOORZHQWU\LQWRD

02'(RURWKHUVSHFLILHGFRQGLWLRQLQWKH$SSOLFDELOLW\ZLWKWKH/&2QRW

PHWDIWHUSHUIRUPDQFHRIDULVNDVVHVVPHQWDGGUHVVLQJLQRSHUDEOH

V\VWHPVDQGFRPSRQHQWVVKRXOGQRWEHDSSOLHGLQWKLVFLUFXPVWDQFH





 $



,IRQH/3&,SXPSLVLQRSHUDEOHWKHLQRSHUDEOHSXPSPXVWEHUHVWRUHGWR

23(5$%/(VWDWXVZLWKLQGD\V,QWKLVFRQGLWLRQWKHUHPDLQLQJ

23(5$%/(SXPSVSURYLGHDGHTXDWHFRUHFRROLQJGXULQJD/2&$

+RZHYHURYHUDOO/3&,UHOLDELOLW\LVUHGXFHGEHFDXVHDVLQJOHIDLOXUHLQ

RQHRIWKHUHPDLQLQJ23(5$%/(/3&,VXEV\VWHPVFRQFXUUHQWZLWKD

/2&$PD\UHVXOWLQWKH/3&,VXEV\VWHPVQRWEHLQJDEOHWRSHUIRUPWKHLU

LQWHQGHGVDIHW\IXQFWLRQ7KHGD\&RPSOHWLRQ7LPHLVEDVHGRQD

UHOLDELOLW\VWXG\FLWHGLQ5HIHUHQFHWKDWHYDOXDWHGWKHLPSDFWRQ(&&6

DYDLODELOLW\DVVXPLQJYDULRXVFRPSRQHQWVDQGVXEV\VWHPVZHUHWDNHQ

RXWRIVHUYLFH7KHUHVXOWVZHUHXVHGWRFDOFXODWHWKHDYHUDJHDYDLODELOLW\

RI(&&6HTXLSPHQWQHHGHGWRPLWLJDWHWKHFRQVHTXHQFHVRID/2&$DV

DIXQFWLRQRIDOORZDEOHUHSDLUWLPHV LH&RPSOHWLRQ7LPHV 



0RQWLFHOOR  % -_I 5HYLVLRQ1R