ML20246P285
| ML20246P285 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/31/1989 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp |
| Shared Package | |
| ML20246P289 | List: |
| References | |
| DPR-50-A-151 NUDOCS 8909110080 | |
| Download: ML20246P285 (8) | |
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i METROPOLITAN EDIS0N COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORPORATION DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPEPATING LICENSE Amendment No.151 License No. OPR-50 1.
The Nuclear Pegulatory Comission (the Commission) has found that:
A.
The application for amendment by GPU Nuclear Corporation, et al.
(the licensee) cated June 13, 1989 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confonnity 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 the Comission's regulations and all applicable requirements have been satisfied.
8909110080 890831 FDR f4 DOCK 05000289 p
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, l 2.
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 Faciiity Operating License No. DPR-50 are hereby amended to read as follows:
(2)' Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.151, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance,' to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 4-s' b0-Mw
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John F. Stolz, Director Project Directorate I-4 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 31, 1989
_____---__ ____---_-__ ~
e, ATTACHMENT TO LICENSE AMENDMENT NO.151 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Replace the following pages of the Facility Operating License:and the' Appendix-A Technical. Specifications with the attached pages. The revised pages are
' identified by amendment number and contain vertical lines indicating the area of change.--
Remove Insert 4-32 4-32 4 4-33 4-34 4-34 4-34a-4-34a 4-34b 4-34b 4-34c
__.__.__________._1_.____._____.____
detection tests.
Sufficient data and analysis shall be included to show that a stabilized leak rate was attained and to identify all significant required correction factors such as those associated with humidity and barometric pressure, and all significant errors such as those' associated with instrumentation sensitivities and data scatter.
This report shall be titled " Reactor Containment Building Integrated Leak Rate Test" and shall be submitted to the NRC within 3 months of the test.
4.4.1.2 Local Leakage Rate Tests 4.4.1.2.1 Scope of Testing Local Leakage Rate tests of penetrations and valves identified in the FSAR shall be performed in accordance with 10CFR 50 Appendix J except as provided in 4.4.1.2.5.f.
4.4.1.2.2 Conduct of Tests
- a. Local leak rate tests shall be performed pneumatically at a pressure of not less than P.,
with the following exception: The access hatch door seal test shall normally be performed at 10 psig and the test every six months specified in 4.4.1.2.5.b shall be performed at a pressure not less than P..
- b. Acceptable methods of testing are halogen gas detection, pressure decay, pneumatic flow measurement, or equivalent.
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- c. The pressure for a valve test shall be applied in the same direction as that when the valve would be required to perfor:n its safety function unless it can be determined that the direction will provide equivalent or more conservative results,
- d. Valves to be tested shall be closed by normal operation and without any preliminary exercising or adjustments.
4.4.1.2.3 Acceptance Criteria
)
The combined leakage from all penetrations and valves subject to Local Leak Rate tests shall not exceed.6 L. (the maximum allowable leakage rate at P.).
4.4.1.2.4 Corrective Action and Retest
- a. If at any time it is determined that the criterion of 4.4.1.2.3 above is exceeded, repairs shall be initiated imediately.
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- b. If conformance to the criterion of 4.4.1.2.3 is not demonstrated within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following detection of excessive local leakage, the reactor shall be shutdown and depressurized until repairs are l
effected and the local leakage meets the acceptance criterion as demonstrated by retest.
4-32 Amendment Nos. pf, Sf, g,151 l
4.4.1.2.5-Test Frequency
+
Local leak detection tests shall be performed at a frequency as required by 10CFR 50 Appendix J, except that:
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- a. The equipment hatch and fuel transfer tube seals shall be tested
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every other refueling period but in no case at intervals greater than 3 years.
If they are opened they will be tested after being closed.
- b. The entire personnel and emergency airlocks shall be tested once every six months. When the airlocks are opened'during the interim between six month tests, the airlock door resilient seals shall ce tested within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the first of each of a series of openings. This requirement exists whenever containment integrity is required.
- c. The recctor building purge isolation valves sha
be leak tested per 10CFR 50, Appendix J, Item III.D.3.
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- d. An. interspace pressurization test (See T.S. 4.4.1.7.1) shall be performed for reactor building purge isolation valves every 3 months. This requirerient is not in effect during cold shutdown.
- e. Cd htad.
- f. Where an exemption from the frequency specified by 10CFR 50 Appendix J has been granted by the NRC, the frequency specified by the exemption shall apply.
4.4.1.3 Isolation Valve Functional Tests Every three months, remotely operated reactor building isolation valves shall be stroked to the position required to fulfill their safety function unless such operation is not practical during plant operation. The valves not stroked every three months shall'be stroked during each refueling period.
4.4.1.4 Annual Inspection A visual examination of the accessible interior and exterior surfaces of the containment structure and its components shall be performed annually and prior to any integrated leak test to uncover any evidence of deterioration which may affect either the containment's structural integrity or leak-tightness. The discovery of any significant deterioration shall be accompanied by corrective actions in accord with acceptable procedures, nondestructive tests, and inspections, and local testing where practical, prior to the conduct of any integrated leak test.
Such repairs shall be reported as part of the test results.
4-33 Amendment Nos. M.6, M, J4fI,151 m
4 4'4.1.5 Reactor Building Modifications Any major modification or replacement of components affecting the reactor building integrity shall be followed by either an integrated leak rate test or a local leak test, as appropriate, i.
and shall meet the acceptance criteria of 4.4.1.1.6 and l
4.4.1.2.3; respectively.
4.4.1.6 Operability.of Access Hatch Interlocks 1.
At least once per six months the operability of the personnel and emergency hatch door interlocks and the associated control room annunciator circuits shall be determined.
If the interlock permits both doors to be open at the same time or does not provide accurate status indication'in the control room the interlock shall be o
declared inoperable.
2.
During periods when containment integrity is required and an interlock-is inoperable, each entry and exit via that airlock shall be locally supervised by a member of the unit operating maintenance or technical staffs, to assure that only one door is open at any time and that both doors are properly closed following use. A record of supervision and verification of closure shall be maintained during periods of-interlock inoperability in an appropriate station log.
3.
If an. interlock is inoperable fer more than 14 days following determination of inoperability, use of the airlock, except for emergency purposes, shall be suspended until the interlock is returned to operable status.
4.4.1.7 Operability of Purge Valves 1.
A periodic pressurization of the purge valve interspaces to 50.6 psig per Specification 4.4.1.2.5.d shall be performed l
to help assure timely detection and resolution of valve and/or actuator degradation.- The acceptance criteria is that total local. leakage when updated for the new purge valve leakage shall be less than 0.6L..
See Specification l
3.6.8 for further action.
2.
The rubber seats on purge valves shall be visually examined each refueling interval to detect degradation (e.g.
cracking, brittleness, etc.) and to assure-timely cleaning, lubrication, and seat replacement. As a minimum, seats shall be replaced at the first refueling-following 5 years of seat service.
4-34 Amendment Nos, f#, MM, M 151 C
-4 f
Bases
(')
The reactor building is. designed for an internal pressure of 55 psig and a steam-air mixture temperature of 281*F.
Prior to initial operation, the containment was strength tested at 115 percent of design pressure and leak rate tested at the design pressure. The containment was also leak tested prior to. initial operation at approximately 50 percent of the design pressure.
These tests established the acceptance criteria of 4.4.1.1.3.
The performance of periodic integrated and local leakage rate tests during the plant life provides a current assessment of potential leakage from the. containment in case of an accident that would pressurize the 'nterior of the containment.
In. order to provide a realistic appraisal of the integrity of the containment under accidert conditions "as found" local leakage results must be i
documented for correction of the integrated leakage rate test results.
Containment isolation valves are to be closed in the normal manner prior to local or integrated leakage rate tests. Containment Isolation Valves are addressed in the FSAR.
The minimum test pressure of 30 psig for the periodic integrated I
i leakage rate test is sufficiently high to provide an accurate measurement of the leakage rate and it exceeds the pre-operational leakage rate test at the reduced pressure of 27.5 psig. The I
specification provides a relationship for relating the measured leakap9 of air at the reduced pressure to the potential leakage of 55 psig. The minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was specified for the integrated leakage rate test to. help stabilize conditions and thus improve accuracy and to better evaluate data scatter.
The frequency of the periodic integrated leakage rate test is keyed to the refueling schedule for the reactor, because these tests can best be performed during refueling shutdowns.
The specified frequency of periodic integrated leakage rate tests is based on three major considerations. First is the low probability of leaks in the liner, because of conformance of the complete containment to a 0.10 percent leakage rate at 55 psig during pre-operational testing and the absence of any significant stresses in the liner during reactor operation.
Second is the more frequent testing, at design pressure, of those portions of the containment envelope that are most likely to develop leaks during reactor operation and the low value of leakage that is specified as acceptable from penetrations and isolation valves, 0.6 L.,
l 4-34a Amendment Nos. g, )6, JHlf, J9ff, Jt( 151
1*
More frequent. testing of various penetrations is specified as these locations are-more susceptible to leakage than.the reactor building liner due to the mechanical closure involved. The basis for specifying a total leakage rate of 0.6 L. from those penetrations l
and isolation valves is that more than one-half of the allowable integrated leakage rate will be from these sources.
. Valve operability tests are specified to assure proper closure or opening of the reactor building isolation valves to provide for isolation or functioning of. Engineered Safety Features systems.
Valves will be stroked to the position required to fulfill their safety function unless it is establish that such testing is not practical during operation. Valves that cannot be full-stroke tested will be part-stroke tested during operation and full-stroke tested during each normal refueling shutdown.
Periodic surveillance of the airlock interlock systems is specified to assure, continued operability and preclude instances where one or both doors are inadvertently left open. When an airlock is inoperable and containment integrity is required, local supervision of airlock operation is specified.
Purge valve interspace pressurization test operability requirements and inspections provide a high degree of assurance of purge valve performance as containment isolation barriers. Third is the tendon stress surveillance program which provides assurance that an important part of the structural integrity of the containment is maintained.
Reference (1) FSAR, Chapter 5.
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l 4-34b AmendmentNos.g,g,pHf,Jf3, 151 m