ML19162A093

Issuance of Amendment No, 202 Regarding Deletion of the Note Associated with Technical Specification 3.5.1., Erccs - Operating

ML19162A093
Person / Time
Site:	Monticello
Issue date:	07/30/2019
From:	Robert Kuntz Plant Licensing Branch III
To:	Church C Northern States Power Company, Minnesota
Kuntz R, NRR/DORL, 415-3733
References
EPID L-2018-LLA-0307
Download: ML19162A093 (13)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 30, 2019 Mr. Christopher Church Senior Vice President and Chief Nuclear Officer Northern States Power Company - Minnesota 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT- ISSUANCE OF AMENDMENT NO. 202 RE: DELETION OF THE NOTE ASSOCIATED WITH TECHNICAL SPECIFICATION 3.5.1, "ECCS -OPERATING" (EPID L-2018-LLA-0307)

Dear Mr. Church:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 202 to Renewed Facility Operating License No. DPR-22, for the Monticello Nuclear Generating Plant. The amendment consists of changes to the Technical Specifications in response to your application dated November 12, 2018, as supplemented by letter dated April 18, 2019.

The amendment deletes a note associated with the limiting condition of operation (LCO) 3.5.1 of Technical Specifications Section 3.5.1, "ECCS [emergency core cooling system] - Operating."

The deleted note permitted the low pressure coolant injection system to be considered operable under certain conditions in Mode 3.

A copy of our related safety evaluation is also enclosed. A Notice of Issuance will be included in the Commission's bi.weekly Federal Register notice.

<bert'F. Ku tz, S~ior Project Manager Plant Licensin Br 'nch Ill Division of Oper ng Reactor Licensing Office of Nuclear eactor Regulation Docket No. 50-263

Enclosure:

1. Amendment No. 202 to DPR-22

2. Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NORTHERN STATES POWER COMPANY- MINNESOTA DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 202 Renewed License No. DPR-22

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A The application for amendment by Northern States Power Company - Minnesota dated November 12, 2018, as supplemented by letter dated April 18, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act}, and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-22 is hereby amended to read as follows:

The Technical Specifications contained in Appendix A and Appendix B, as revised through Amendment No. 202, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of the date of issuance.

Lisa M. Regner, Acting Chief Plant Licensing Branch 111 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: July 30, 2019

ATTACHMENT TO LICENSE AMENDMENT NO. 202 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE DOCKET NO. 50-263 Renewed Facility Operating License No. DPR-22 Replace the following page of the Renewed Facility Operating License No. DPR-22 with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating area of change INSERT REMOVE Page 3 Page 3 Technical Specifications Replace the following pages of Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

INSERT REMOVE 3.5.1-1 3.5.1-1

2. Pursuant to the Act and 10 CFR Part 70, NSPM to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operations, as described in the Final Safety Analysis Report, as supplemented and amended, and the licensee's filings dated August 16, 1974 (those portions dealing with handling of reactor fuel);

3. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required;

4. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and

5. Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess, but not separate, such byproduct and special nuclear material as may be produced by operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission, now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

1. Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 2004 megawatts {thermal).

2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 202, are hereby incorporated in the license. NSPM shall operate the fa~ility in accordance with the Technical Specifications.

3. Physical Protection NSPM shall implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Renewed License No. DPR-22 Amendment No. 202

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 pressures 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 One Core Spray subsystem inoperable.

Monticello 3.5.1-1 Amendment No. 202 I

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 202 TO RENEWEDFACILITY OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY- MINNESOTA MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263

1.0 INTRODUCTION

By application dated November 12, 2018 (Agencywide Documents Access and Management System (ADAMS Accession No. ML18317A172), as supplemented by letter dated April 18, 2019 (ADAMS Accession No. ML19108A223), Northern States Power Company- Minnesota, doing business as Xcel Energy (NSPM, the licensee) requested changes to the technical specifications {TSs) for Monticello Nuclear Generating Plant (MNGP). The supplemental letter dated November 12, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register on January 2, 2019 (84 FR 24).

The proposed change would revise TS 3.5.1, "ECCS [emergency core cooling system] -

Operating," by erasing the note in limiting condition of operation (LCO) 3.5.1.

2.0 REGULATORY EVALUATION

2.1 Plant and System Description MNGP is a single cycle, forced circulation, low power density boiling-water reactor (BWR)/3 with a Mark I pressure suppression type primary containment. As described in Section 5.1 of the Updated Safety Analysis Report (USAR), the primary containment encloses the reactor pressure vessel (RPV), the reactor coolant recirculation loops, and other branch connections of the reactor coolant system (RCS). The major elements of the primary containment are the drywall, the pressure suppression chamber (or wetwell) that stores a large volume of water (suppression pool), the connecting vent pipe system between the drywell and the wetwell, isolation valves, the vacuum relief system, the containment cooling systems and service equipment.

Enclosure 2

MNGP USAR Section 6.2.3, "Residual Heat Removal System (RHR)," and Section 10.2.4, "Reactor Shutdown Cooling System," describes the RHR system, its functions, and operating modes. The RHR system consists of two subsystems. Each subsystem consists of a heat exchanger, two pumps in parallel, and associated piping, and 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 for either RHR subsystem to supply water to the other subsystem.

Among the RHR system operating modes, the two modes that are related to the proposed license amendment request are the LPCI mode and the shutdown cooling (SDC) mode. Each RHR subsystem aligned in the LPCI mode consists of a suppression pool suction line, parallel flow paths through the two pumps, and a common injection line to the RPV. The SDC subsystem consists of a suction line from one of the reactor recirculation loops, parallel flow paths through the pump(s), and heat exchanger to discharge heat to the RHR service water system, and then return to the RPV through either one of the recirculation loops.

The RHR system in the LPCI mode is a part of the ECCS. Besides this mode, the ECCS network consists of the HPCI system, the core spray (CS) system, and the automatic depressurization system. The LPCI subsystem restores and maintains RPV inventory for adequate core cooling during a LOCA at low RPV pressure.

The RHR system in its SDC mode is used to remove decay and residual heat from the reactor during a normal reactor shutdown and cooldown and lower the RCS inventory temperature below 212 degrees Fahrenheit (°F) prior to servicing and refueling operations. The SDC subsystem is placed in service in Mode 3 (hot shutdown) when the reactor steam dome pressure is less than the SDC supply isolation valve interlock pressure. This interlock is provided to prevent the over-pressurization of the RHR SDC suction line by inadvertent operation of the suction isolation valve from the recirculation loop. Operation in SDC is not permitted when the reactor steam dome pressure is greater than the pressure interlock setting of the SDC suction isolation valve.

2.2 Proposed Change The change proposed in the license amendment request would revise LCO 3.5.1 of TS 3.5.1, to delete the following 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.

2.3 Regulatory Requirements The following regulatory requirements are applicable to the proposed amendment:

• Title 10 of the Code of Federal Regulations Section 50.36, "Technical Specifications,"

details the content and information that must be included in a plant TS. It requires to include (1) safety limits, limiting safety system settings, and limiting control settings; (2)

LCOs; (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. As described in 10 CFR 50.36(c)(2), the LCOs 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 remedial action permitted by the technical specifications until the condition can be met.

• Section 50.46(a)(1 )(i) of 10 CFR 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 loss-of-coolant accident (LOCA).

2.4 Applicable General Design Criteria In 1967, the Atomic Energy Commission (AEC) published for comments an amendment to 10 CFR 50, Appendix A, a revised set of proposed General Design Criteria (GDC) in the Federal Register (32 FR 10213), dated July 11, 1967. The purpose of the amendment was to provide guidance to applicants to develop the Principal Design Criteria (PDC) to be Included in applications for construction permits. The PDCs used for the MNGP design basis are listed in USAR Section 1.2, "Principal Design Criteria". An evaluation which compares the MNGP design basis PDC to the AEC proposed GDC of 1967 is presented in MNGP USAR, Appendix E "Plant Comparative Evaluation with the Proposed AEC 70 Design Criteria." While MNGP is not explicitly licensed to the 1967 AEC proposed GDCs or the current GDCs, the following AEC proposed GDCs and the current GDCs equivalent to the MNGP PDC are applicable to the proposed amendment:

AEC Proposed GDC 6, "Reactor Core Design (Category A}," states that:

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.

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

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.

AEC proposed GDC 44, "Emergency core cooling capability (Category A)," states that:

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.

The corresponding 10 CFR 50, Appendix A, GOC 35, "Emergency core cooling," states that:

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 ....

3.0 TECHNICAL EVALUATION

The note proposed for deletion in TS LCO 3.5.1 states that the RHR system in its LPCI mode may be considered operable during alignment and operation for decay heat removal with reactor steam dome pressure less than the RHR SOC isolation interlock in Mode 3, if capable of being manually realigned and not otherwise inoperable. The license amendment request states the following issue in the RHR system operation as the reasons for deletion of this note:

... 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 lnoperability", dated June 16, 2010. This phenomenon is applicable to some boiling water reactor plants due to the physical arrangement (that is the common interface) of the SOC and LPCI suction lines for the RHR pumps. Realignment from SOC to the LPCI mode transfers the suction source for the RHR pump. The resultant flashing or boiling of the high pressure / high temperature SOC 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 SOC water temperature is the highest.

NSPM does not have an analysis to demonstrate that realignment of an RHR subsystem from SOC 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.

In the TS Bases for LCO 3.5.1, it is stated that all ECCS subsystems are required to be operable during Mode 3, when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a LOCA. In Mode 3, when reactor steam dome pressure is 150 pounds per square inch gauge (psig), automatic depressurization

system and HPCI are not required to be operable because the low pressure ECCS subsystems, which include the RHR LPCI subsystem and the CS system, can provide sufficient flow below this pressure.

USAR Section 10.2.4.3 states that:

The reactor shutdown cooling system is placed into operation during a normal plant cooldown when reactor dome pressure is below 81.8 psig. Operation of this portion of the RHR system for shutdown cooling does not compromise the ability of the RHR system to operate in the low-pressure coolant injection system (LPCI) mode. During shutdown, the probability of requiring LPCI operation is very low. However, if LPCI operation is required, the operator can manually terminate shutdown cooling and start LPCI operation from the main control room.

The above USAR statement allows the RHR system be transferred to the LPCI mode of RHR system from the main control room while in SOC operation, if required. The transfer to LPCI mode may be completed in the event of a LOCA in Mode 3 when the reactor water has cooled enough to avoid the possibility of water/steam hammer in the RHR pump suction piping. The April 18, 2019 supplement stated that the alignment to the LPCI mode is still permitted, if required, in the region between the RHR shutdown cooling supply isolation interlock interlock and cold shutdown/ refueling conditions. To achieve LPCI alignment, if required, would require both control room and outside control room operations in accordance with established plant procedures.

The SOC mode can be initiated in Mode 3 at RPV pressure below 81.8 psig using both RHR subsystems. In this scenario both RHR subsystems would be declared inoperable in the LPCI mode because of the identified potential for voiding in the pump suction piping, RHR pump cavitation, water hammer, and potential RHR system damage. Since LCO 3.5.1 is applicable in Mode 3 upon initiation of SOC using both RHR subsystems, the plant could be in a situation where LCO 3.5.1 was not met for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> according to TS 3.5.1 Condition D. As stated in USAR Section 10.2.4.2, the RHR SOC mode can complete cooldown to 125 °F in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and maintain the nuclear system at 125 °F or less. Therefore, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the initiation of SOC, the LCO would no longer be applicable, the reactor would be in Mode 4, and LPCI subsystems may be re-aligned from the SOC mode if required, without the potential for voiding in the pump suction piping, RHR pump cavitation, water hammer, and RHR system damage.

The NRC staff review confirmed that the potential for 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 SOC mode to LPCI mode.

Based on the above, the NRC staff finds that the current note in LCO 3.5.1 could potentially allow operating conditions to exist that could adversely impa_ct the function of the RHR system because high pressure, high temperature water when introduced to the low-pressure piping could result in voiding in the suction piping, RHR pump cavitation, water hammer, and associated RHR system damage. Therefore, the NRC staff finds that removal of the note in TS LCO 3.5.1 is acceptable and conservative, and the applicable regulatory requirements will continue to be met.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendment on date June 10, 2019. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (84 FR 24). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that 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.

Principal Contributor: Ahsan Sallman, NRR Dateofissuance: July 30, 2019

ML19162A093 *memo date **via e-mail OFFICE NRR/DORL/LPL3/PM N RR/DORL/LPL3/LA NRR/DSS/SRXB/BC N RR/DSS/STSB/BC(A)

NAME RKuntz SRohrer JWhitman* PSnyder **

DATE 6/13/19 6/13/19 5/20/19 7/2/19 OFFICE OGC NLO NRR/DORL/LPL3/BC(A) NRR/DORL/LPL3/PM NAME DRoth** LRegner (MOrenak for) RKuntz DATE 7/22/19 7/30/19 7/30/19