ML20059H831
| ML20059H831 | |
| Person / Time | |
|---|---|
| Site: | Millstone  (DPR-21-A-067, DPR-21-A-67) |
| Issue date: | 11/01/1993 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059H743 | List: |
| References | |
| NUDOCS 9311100238 | |
| Download: ML20059H831 (9) | |
Text
{{#Wiki_filter:_ c0 * "%<<b ~ j" .S *' ( E UNITED STATES i 7 %C NUCLEAR REGULATORY COMMISSION g '.w...,# W ASHINGTON, D.C. 205S4001 8 NORTHEAST NUCLEAR ENERGY COMPANY DOCKET NO. 50-245 MILLSTONE NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 67 License No. DPR-21 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Northeast Nuclear Energy Company (the licensee), dated July 27, 1993, 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 4 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. 1 9311100238 931101 PDR ADOCK 05000245 1 P PDR )
I 2. A. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraphs 2.C.(2) of Facility Operating License No. DPR-21 is hereby amended to read as follows: 2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 67 are hereby incorporated in the license. l The licensee shall operate the facility in accordance with the 1 Technical Specifications. B. Facility Operating License No. DPR-21 is further amended by revising paragraph 2.D(2) to read as follows: (2) Appendix J to 10 CFR Part 50, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." Appendix J relates to containment leakage test requirements, specifically periodic verification by tests of the leak-tight integrity of the primary reactor containment and systems and components which penetrate containment. Three exemptions were granted on May 10, 1985, on the basis of the staff safety evaluation. The two which remain in effect relate to testing of expansion bellows at containment penetrations and main steam isolation valves. 3. This license amendment is effective as of the date of issuance, to be implemented within 60 days of issuance. l FOR THE NUCLEAR REGULATORY COMMISSION n \\ .ohn h stolz, irector (Pro,J' ct Directorate I-3 B Wision of Reactor Projects - I/II Office of Nuclear Reactor Regulation Attachments: 1. Pages 4 and 5 of License
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Changes to the Technical Specifications Date of Issuance: November 1, 1993 Pages 4 and 5 are attached, for convenience, for the composite license to reflect this change. Page 3a can be discarded. i o
4 (5) Intearated Imolementation Schedule a. Northeast Nuclear Energy Company shall implement and maintain in effect the Integrated Implementation Schedule Program Plan (the Program Plan) to be followed for scheduling of plant modifications and engineering studies. The Program Plan shall be followed from and after the effective date of this license condition. b. This license condition shall be effective for three years from the date of issuance of Amendment No. 56. D. The facility has been granted certain exemptions from the requirements of 10 CFR Part 50 as set forth below: ) (1) Section III.G. of Appendix R to 10 CFR Part 50, " Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979." This section relates to fire protection features for ensuring the systems and associated circuits used to achieve and maintain safe shutdown are free of fire damage. The staff safety evaluation, dated November 6,1985, concluded that the licensee's existing fire-protection configuration with proposed modifications achieves an equivalent level of safety. Exemption granted November 6, 1985. (2) Appendix J to 10 CFR Part 50, " Primary Reactor Containment l Leakage Testing for Water-Cooled Power Reactors." Appendix J relates to containment leakage test requirements, specifically l periodic verification by tests of the leak-tight integrity of l the primary reactor containment and systems and components which penetrate containment. Three exemptions were granted on May 10, 1985, on the basis of the staff safety evaluation. The two which remain in effect relate to testing of expansion bellows at containment penetrations and main steam isolation valves. (3) Section 50.71 of 10 CFR Part 50, " Maintenance of Records, Making of Reports." Section 50.71(e)(3) relates to the j requirement for submittal of an updated FSAR. An exemption for a schedular delay in the submittal of the updated FSAR was granted on November 22, 1985. This exemption requires that the FSAR update be completed and submitted by March 31, 1987. Amendment No. 56, 67
f 5 l 3. This license is effective as of its date of issuance and shall expire at midnight, October 6, 2010. FOR THE NUCLEAR REGULATORY COMMISSION Original signed by: Frank J. Miraglia, Director Division of PWR Licensing - B
Attachment:
Appendix A - Technical Specifications Date of Issuance: October 31, 1986. Amendment No. Is,67 Correction Letter.9/22/88
l l ATTACHMENT TO LICENSE AMENDMENT NO. 67 FACILITY OPERATING LICENSE NO. DPR-21 DOCKET NO. 50-245 j Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Remove Insert i 3/4 7-5 3/4 7-5 .l 1 3/4 7-Sa B 3/4 7-9 8 3/4 7-9 B 3/4 7-9a F k I 4 I
SURVEILLANCE REOUIREMENTS (Continued) l 4.7 CONTAINMENT SYSTEMS i 4.7.A.3.a.4. All test leakage rates shall be calculated using observed data converted to absolute values. Error analyses shall I be performed to select a balanced integrated leakage i measurements system. b. Acceptance Criteria for IPCLT 1. The maximum allowable leak rate at P, air per 24 hours). shall not exceed L (1.2 weight percent of the contained l 2. The allowable operational leak rate, L which shall be metpriortoincreasingreactorcoolankg,ystemtemperature s above 212*F following a test (either as measured or following repairs and retest), shall not exceed 0.75 L,. c. Corrective Action for IPCLT If leak repairs are necessary to meet the allowable operational leak rate, the integrated leak rate test need not be repeated provided local leak measurements are conducted and the leak rate differences prior to and after repairs, when corrected to P and deducted from the integrated leak rate measurements, yield a leakage rate value not in excess of the allowable operational leak rate Lto-d. Local Leak Rate Tests (LLRT) i (1) Primary containment testable penetrations and isolation valves shall be tested at a pressure of 43 psig, except the main steam line isolation valves shall be tested at a pressure of 25 psig each operating cycle. Bolted double-gasketed seals shall be tested whenever the seal is closed after being opened and at least once during each operating cycle. (2) Personnel Air Lock Door Seals i) The personnel air lock door seals shall be tested at a pressure of 43 psig at least once every six months. ii) Prior to entering a mode in which primary containment integrity is required, the personnel air lock door seals shall be tested at 43 psig if the air lock was opened during the period when primary containment integrity was not required. iii) The personnel air lock door seals shall be tested at 43 psig within 72 hours following any opening of the air lock during periods when primary containment integrity is required. Millstone Unit 1 3/4 7-5 Amendment No. 67 l 0102 i
9 SURVEf tt ANCE REOUIREMENTS (Continued) i j
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e. Acceptance criteria and corrective action for LLRT: -l If the total leakage rates listed below are exceeded, repairs and retests shall be performed to correct the condition. i .l l l -{ i i a i i I i .. i i I i
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l l'l] .i l l i Millstone Unit 1 3/4 7-Sa Amendment No. 67 l j nio I ] .. ~. :
I i 4.7 CONTAINMENT SYSTEMS BASES The penetration and air purge piping leakage test frequency, along with the containment leak test, which is performed at a test pressure of at least 43 psig (P Wheneveradoubl8-)g,asketedpenetration(primarycontainmenthead,is adequate equipment hatches, and the suppression chamber access hatch) is broken and remade, tne space between the gaskets is pressurized to determine that the seals are performing properly. The test pressure of at least 43 maximum l psig is ccnsistent with the accident analyses and the preoperational leak rate test pressure. It is expected that the majority of the leakage from valves, penetrations and seals would be into the reactor building. However, it is possible that leakage into other parts of the facility could occur. Such leakage paths that may affect significantly the consequences of accidents are to be minimized. Personnel air lock door seal testing is performed in accordance with f 10CFR50, Appendix J requirements. Monitoring the nitrogen makeup requirements of the inerting system provides a method of observing leak rate trends and would detect gross leaks in a very short time. This equipment must be periodically removed from service for test and maintenance, but this out-of-service time will be kept to a practical minimum. Surveillance of the suppression chamber-drywell vacuum breakers consists of operability checks, calibration of instrumentation and inspection of the valves. The monthly operability tests are performed to check the capability of the disc to open and close and to functionally test the position indication system. This test frequency is justified based on previous experience and the fact that these valves are normally closed and are only open during tests or accident conditions. t The refueling outage surveillance tests are performed to check that the valve will perform properly during the accident condition and to verify the calibration of the position indication system. Measuring the force required to lift the valve assures that the valve will function properly during an accident. Inspection of a select number of valves during each refueling outage assures that deterioration of the valve internals or misalignment of the disc does not impair the proper operation of the valve. j This test interval is based on equipment quality and previous equipment experience. \\ i l Millstone Unit 1 B 3/4 7-9 Amendment No. JJ, 67 olos ~
4.7 CONTAINMENT SYSTEMS BASES i B. Standby Gas Treatment System and 4 C. Secondary Containment Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 7 inches of water, at the system design flow rate, will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Heater capability and pressure drop should be determined at least once per operating cycle to show system performance capability. i i I l l 4 e a i i i !i i i Millstone Unit 1 B 3/4 7-9a Amendment No. Jijg, 6? 0103 o ? _ _ _ _ _ _ _ _ _, _, _ _ _ _.}}