ML20101N160

Amend 95 to License DPR-22 Modifying TS to Revise Main Sliv Leak Rate Test Acceptance Criterion,Operability Test Interval for Drywell Spray Header & Nozzles from 5 to 10 Yrs & TS 3/4.7.a.2

ML20101N160
Person / Time
Site:	Monticello
Issue date:	04/03/1996
From:	Kim T NRC (Affiliation Not Assigned)
To:
Shared Package
ML20101N163	List: ML20101N160
References
NUDOCS 9604080276
Download: ML20101N160 (9)

Text

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t UNITED STATES g

g NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 o

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NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 95 License No. DPR-22 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Northern States Fower Company (the licensee) dated August 15, 1995, as supplemented November 14, and December 20, 1995, 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 Comission; 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 Comission'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 I

the Comission's regulations and all applicable requirements have been j

satisfied.

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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 Facility Operating License No. DPR-22 is hereby amended to read as follows:

9604080276 960403 PDR ADOCK 05000263 P

PDR

Technical Snecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 95

, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMISSION dl Tae Kim, Project Manager Project Directorate III-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: April 3, 1996 l

l

ATTACHMENT TO LICENSE AMENDMENT NO. 95 FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

REMOVE INSERT 104 104 158 158 159 159 160 160 161 184 184 185 l

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3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS C.

Containment Spray / Cooling System C. Containment Spray / Cooling System 1.

Except as specified in 3.5.C.2, 3 and 4 below, 1.

Demonstrate the RHR Service Water pumps both Containment Spray / Cooling Subsystems develop 3,500 gpm flow rate against a 500 shall be operable whenever irradiated fuel is ft head when tested pursuant to in the reactor vessel and reactor water Specification 4.15.B.

temperature is greater than 212*F. A containment / spray cooling subsys. tem consists 2.

Test the valves in accordance with of the following equipment powered from one Specification 4.15.B.

division:

3.

Demonstrate the operability of the drywell 2 RHR Service Water Pumps spray headers and nozzles with an air test 1 Heat Exchanger during each 10 year period.

l 2 RHR Pumps

• Valves and piping necessary for:

Torus Cooling Drywell Spray 2.

One RER Service Water Pump may be inoperable for 30 days.

3.

One RHR Service Water Pump in each subsystem may be inoperable for 7 days.

4.

One Containment Spray / Cooling Subsystem may be inoperable for 7 days.

5.

If the requirements of 3.5.C.1, 2, 3 and 4 cannot be met, an orderly shutdown of the reactor will be initiated and the reactor water temperature shall be reduced to less than 212*F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• For allowed out of service times for the RHR pumps see Section 3.5.A.

3.5/4.5 104 AMENDMENT 27.77.79.95

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 2.

Primary Containment Integrity 2.

Primary Containment Integrity a.

Primary Containment Integrity as defined a.

Primary Containment Integrity shall in Section 1, shall be maintained at all be demonstrated after each closing times when the reactor is critical or of each penetration subject to Type B when the reactor water temperature is testing, If opened following a Type A above 212*F and fuel is in the reactor or Type B test, by leak rate testing vessel, except when performing low power the seal with gas at 2 Pa, 42 psig, physics test at atmospheric pressure and verifying that when the measured during or after refueling at power levels leakage rate for these seals is added not to exceed 5 MW(t). Without Primary to the leakage rates determined pursuant Containment Integrity, restore Primary to Surveillance Requirements 4.7.A.2.b l

Containment Integrity within one hour or for all other Type B and C penetrations, be in at least Hot Shutdown within the the combined leakage rate is less than next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Cold Shutdown within or equal to 0.6La.

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

3.7/4.7 158 AMENDMENT NO. 30,55,60,95

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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS b.

When Primary Containment Integrity is required, b.

Perform required visual examinations and leakage rates shall be limited to:

leakage rate testing for Type A containment integrated leakage rate tests in accordance 1.

An overall integrated leakage rate of less with 10 CFR 50, Appendix J, Option B, as l

than or equal to La, 1.2 percent by weight modified by approved exemptions, and Regulatory of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa, Guide 1.163 dated September 1995. Perform Type 42 psig.

B and C tests in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved 2.

A combined maximum flow path leakage rate exemptions.

of less than or equal to 0.6La for all penetrations and valves, subject to Type B 1.

Deleted and C tests when pressurized to Pa, 42 psig.

2.

Deleted 3.

Lass than or equal to 46 scf per hour 3.

Deleted combined maximum flow path leakage for all main steam isolation valves when tested at 4.

Deleted 25 psig.

5.

Deleted With the measured overall integrated primary containment leakage rate exceeding 0.75La, or the measured combined leakage rate for all penetrptions and valves subject to Type B and C testing exceeding 0.6La, or the measured combined maximum flow path leakage rate exceeding 46 scf per hour for all main steam isolation valves, restore leakage rates to less than or equal to these values prior to increasing reactor coolant system temperature above 212*F or, alternatively, restore measured 1

leakage rates to within these limits within one hour or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown within the followinc 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

159 AMENDMENT NO. 52,53,72,95

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i 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS l

c.

Whrsn Primary Contairmaent Integrity is required, c.

The primary containment airlock shall be the primary containment airlock shall be demonstrated operable:

operable with:

1.

By performing overall airlock leakage rate 1.

Both doors closed except when the airlock testing in accordance with 10 CHL 50, is being used, then at least one airlock Appendix J, as modified by approved door shall be closed, and exemptions.

2.

An overall airlock leakage rate of less 2.

Deleted than or equal to 0.05La at Pa or 0.00714 at 10 psig.

3.

At six month intervals by verifying that only one door can be opened at With the primary containment airlock a time.

If the airlock has not been inoperable, maintain at least one airlock door used since the last door interlock closed and restore the airlock to Operable test, this test is not required.

status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold d.

The interior surfaces of the drywell shall Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

be visually inspected each operating cycle for evidence of deterioration.

3.7/4.7 160 AMENDMENT NO. 55, 60, 95

Bases Continued:

While the design of the Monticello plant predates 10 CFR Part 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," testing substantially conforms to the requirements of Appendix J.

The design of the plant was thoroughly reviewed to determine where compliance with Appendix J was impossible or impractical.

In each case where a departure from the requirements of Appendix J was identified, a request for exemption from the requirements of Appendix J or a plant modification was proposed and submitted for NRC Staff review. Exemptions were proposed in those cases where compliance with Appendix J would have provided no meaningful improvement in plant safety.

4 In their review of Appendix J compliancem, the NRC Staff approved a number of exemption requests, denied others, and provided necessary interpretation and clarification of the requirements of Appendix J.

The Technical Specification surveillance requirements reflect the results of this

review, a

Exemption from the requirements of Appendix J was provided in the following areas:

i a.

Testing of valves sealed by water b.

Low pressure testing of main steam line isolation valves c.

Low pressure testing of the primary containment airlock d.

Reduced airlock testing frequency when the airlock is in frequent use The Monticello airlock is tested by pressurizing the space between the inner and outer doors.

Individual door seal leakage tests cannot be performed. Since the inner door is designed to seat with containment pressure forcing the door closed, special bracing must be installed for each leakage test.

The outer door must be opened to install and remove this bracing. Because of the complexity of this operation, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> may be necessary to perform a leakage test.

(1) 14tter from D G Eisenhut, Director, Division of Licensing, USNRC, dated June 3,1984, " Safety Evaluation by the Office of NRR, Appendix J Review".

4.7 BASES 184 AMENDMENT No. 55, 95

Bases Continued:

l On September 26, 1995, Regulatory Guide 1.163 became effective providing guidance on performance based testing to the requirements of 10 CFR 50, Appendix J, Option B.

Monticello has adopted Option B,Section III.A of 10 CFR Part 50, Appendix J, for Type A primary reactor containment integrated leakage rate testing. Monticello will continue to perform Type B and C testing in accordance with 10 CFR Part 50, Appendix J, Option A.

.r 4.7 BASES NEXT PAGE IS PAGE 188 185 AMENDMENT NO. 95

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