ML19350D199
| ML19350D199 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/30/1981 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co |
| Shared Package | |
| ML19350D200 | List: |
| References | |
| DPR-50-A-063 NUDOCS 8104130778 | |
| Download: ML19350D199 (9) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION j
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WASHINGTON, D. C. 20556 U
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METROPOLITAN EDIS0N COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 63 License No. DPR-50 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company (the licensee), dated September 17, 1975 and revised by letters dated October 29, 1975, February 18, 1977 and May 13, 1980 and staff discussions, 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 conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized j
by this amendment can be conducted without endangering the healtn and safety of the public, and (ii) tnat such activities will be l
conducted in compliance with the Commission's regulations; i
D.
The issuance of this amendment will not be inimical to the comon l
defense and security or to the health and safety of the public; and E. f The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements I
have been satisfied.
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2.
Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-50 is I
.hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 63, 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 its issuance.
FOR THE NUCLEAR RE ULATORY COMfilSSION L
Jo w F. Stolz, Chief O rating Reactors Branch #4 ivision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: March 30,1981 4
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dTTACHMENT TO LICENSE AMENDMENT NO. 63 FACILITY OPERATING LICENSE N0, DPR-50 DOCKET N0. 50-289 Revise Appendix A as follows:
Remove Pages Insert-Pages 4-29 4-29 4-31 thru 4-34a 4-31 thru 4-34a The changed areas on the revised pages are shown by marginal lines.
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4.4 REACTOR BUILDING 4.4.1 CONTAINMENT LEAKAGE TESTS Applicability Applies to containment leakage.
Objective To verify that leakage from the reactor building is maintained within allowable limits.
Specification 4.4.1.1 Integrated Leakage Rate Tests 4.4.1.1.1 Design Pressure Leakage Rate The design integrated leakage rate, (L ), fr m the reactor building at the 55 psig d
design pressure, P, is.1 weight percent of the building atmosphere at that d
pressure per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.1.1.2 Allowable Integrated Leakage Rate The maximum allowable integrated leakage rate, (L ), from the reactor building at the calculated peak reactor building internal pre,ssure of 50.6 psig (P,) associated with the design basis accident, shall not exceed.1 weight percent of the building atmosphere at that pressure per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4. 4.1.1. 3 Testing at Reduced Pressure The governing criteria for the periodic integrated leakage rate tests to be performed at the reduced test pressure, Pt (at 30 psig), is the maximum allowable containment test leakage rate, L
- L is equal to 0.077 t
t weight percent of the building atmosphere per 24 hrs.
Amendment No. 63 4-29
4.4.1.1.5 Frequency of Test After the initial pre-operational leakage rate test, two integrated leakage rate tests shall be performed at approximately equal intervals between each major shutdown for inservice inspection to be performed at 10 year intervals.
In addition, an integrated test shall be performed at each 10 year interval, coinciding with the inservice inspection shutdown.
The test shall coincide with a shutdown for major fuel reloading.
4.4.1.1.6 Acceptance Criteria a.
Initial and periodic integrated leakage rate test at P.
t Ltm shall be legs than.75 L -
t b.
If the initial and periodic integrated leakage rate test fails to meet the acceptance criteria of 4.4.1.1.6a, the test schedule applicable to subsequent tests shall be subject to review and approval by the Commission.
If two consecutive periodic integrated leakage rate tests f ail to c.
meet the acceptance criteria of 4.4.1.1.6a, a test shall be performed at each plant shutdown for refueling or every 18 months, whicsever occurs first, until two consecutive tests meet the criteria o f 4.4.1.1.6a.
4.4.1.1.7 Corrective Action and Retest If, during an integrated or supplemental leak rate test, potentially excessive leakage paths are identified which would result in the integrated leak test not meeting the acceptance criteria:
a.
terminate the integrated or supplemental leak rate test, b.
measure the subject leakage using local leakage testing methods, make repairs and/or adjustment, c.
d.
run an integrated leakage rate test.
If the test data from a completed leakage rate test does not meet the acceptance criteria, the integrated leakage rate test need not be repeated provided local leakage rate measurements are made at pressure Pt before and after repair to demonstrate that the leakagt rate reduction achieved by the repairs reduces the overall measured integrated leakage rate to an acceptable value.
4.4.1.1.8 Report of Test Results Each integrated leak rate test will be the subject of a summary technical report which will include a description of test methods used and a summary of local leak 4-31 Amendment No.63
detection tests. Suf ficient 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 instrumenta-tion sensitivities and data scatter. This report shall be titled Reactor Containment Building Integrated Leak Rate Test and shall be submitted to the AEC within 3 months of the test.
- 4. 4.1. 2 Local Leakage Rate Tests 4.4.1.2.1 Scope of Testing The local leak rate shall be measured for the following components a.
using a type "B" test as defined in 10CFR50, Appendix J.
1.
Personnel air lock door gaskets 2.
Emergency air lock door gaskets 3.
The resilient seals on the equipment hatch and fuel transfer tube blind flanges 4.
Reactor Building Purge valves (AH-VlA, B, C, and D) 5.
Blind flanges on both ends of pipe through the following penetrations:
S.1 No. 104 (S/G drains) i S.2 No.105 (S/G cleaning)
S.3 No. 106 (S/G cleaning)
I S.4 No. 210 (S/G annulus drains) s.5 No. 211 (S/G annulus drains) b.
The local leak rate shall be measured for the following isolation valves using a type "C" test as defined in 10CFR50, Appendix J.
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Containment air sample (CM-V1, 2, 3, and 4) 2.
Hydrogen purge discharge system (HP-V1 and V6) 3.
Make-up and Purification (MU-V18, MU-V20, MU-V116, MU-V2A/B, MU-V25) 4.
Industrial Cooler System (RB-V2* and RB-V7) 5.
Core Flood (CF-V12A and B, CF-Vl9A and B, CF-V20A and B) 6.
Nuclear Service Closed Cooling (NS-V35) 7.
Intermediate Closed Cooling (IC-V2) 8.
Sample Valves (CA-V1, CA-V3, CA-V4A/B, CA-V13) 9.
Drain Valves (WDG-V3, WDL-V303 and WDL-V534) 10.
Chemical Addition (CA-192) 4-32 Amendment No.63 i
11.
Nitronen Rupply (NI-V27) 17.
Decay Heat Removal (OH-V69 & V64)
The 'ollowing isolation valves will be tested by testing the c.
Fluid Block System.
Nuclear Service Closed Cooling Water (NS-V4 and NS-V15) 1.
2.
Intermediate Cooling Water (IC-V3, V4 and V6) 3.
Spent Fuel Cooling (SF-V23) 4.
Make-up and Purification (MU-V3 and MU-V26) 5 Reclaimed Water (CA-V189) 6.
Sample Valves (CA-V5A&B and CA-V2) 7.
Drain Valves (WDL-V304, WDG-V4 and WDL-V535)
The following isolation valves or blank flanges will be tested by d.
testing the Penetrative Pressurization System.
1.
Instrument Air (IA-V6 and 1A-V20) 2.
Service Air (SA-V2 and SA-V3) 3.
Leak rate system (LR-V1, 2, 3, 4, 5, 6, 10, and 49)
Blank flanges on Penetrations 414, 415, 416 4.
Incore Inst. Transfer Tube - Bladk flange on Penetration 241 4.4.1.2.2 Conduct of Tests Local leak rate tests shall be performed pneumatically at a pressure a.
of not less than P, with the following exception: The access hatch door seal test sha$1 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
- a Acceptable methods of testing are halogen gas detection, pressure decay, b.
pneumatic flow measurement or equivalent.
The pressure for a valve test shall be applied in the some direction c.
as that when the valve would be required to perform its safety function unless it can be determined that the direction will provide equivalent or more conservative results.
i Valves to he tested shall be closed by normal operation and without d.
any preliminary exercising or adjustments.
4.4.1.2.3 Acceptance Criteria The combined leakage from all items listed in 4.4.1.2.1, except leakage from those valves or devices sealed by the Fluid Block System or Penetration Pressurization System, shall not exceed.6 L, (the maximum allowable leakage rate at P,).
Amendment No. J6, 63 4-33
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4.4.1.2.4 Corrective Action and Retest If at any time it is determined that the criterion of 4.4.1.2.3 a.
above is exceeded, repairs shall be initia'ted immediately, b.
If conformance to the criterion of 4.4.1.2.3 is not demonstated 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 loc t1 leakage, the reactor shall be shutdown and depressurized until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.
4.4.1.2.5 Test Frequency Local leak detection tests shall be performed at a frequency of at least each refueling period, except that:
The equipment hatch and fuel transfer tube seals shall be tested a.
every other refueling period but in no case at intervals greater than 3 years. If they are opened they will be tested af ter bef ug closed.
b.
TheentirepersonneIandemergencyairlocksshallbetestedonceeverysix t
months. When the airlocks are, opened during the. interim between six month tests, the airlock door restilent seals shall be 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 serie's of openings. This requirement exists when-ever cor.tainment integrity is required.
The reactor building purge isolation valves shall be tested yearly.
c.
d.
Readings of the rotameters in each manifold of the penetration pressurization system shall be recorded at periodic intervals not to exceed three months.
4.4.1. 3 Isolation Valve Functional Tests I
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 l
A visual examination of the accessible interior and exterior surfaces of the i
containment structure and its components shall be performed annually and prior to any integrated leak test to uncover any evidence of deterioratJon 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 inspec-l tions, and local testing where practical, prior to the conduct of any integrated l
leak test. Such repairs shall be reported as part of the test results.
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4-34 Amendment No,63
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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, and shall meet the acceptance criteria of 4.4.1.1.5 and 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 declared inoperable.
2.
During periods when containment integrity is required and an interlock in 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 or.e door is open at any time and that both doors are properly closed following use. A record of super-vision and verification of closure shall be maintained during periods of interlock inoperability in an appropriate station log.
3.
If an interlock is inoperable for more than 14 days following deter-mination of inoperability, use of the hirlock, except for emergency purposes, shall be suspended until the interlock is returned to operable status.
Bases (1)
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 These tests established the acceptance criteria of 4.4.1.1.3.
l pressure.
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 interior of the containment.
In order to provide a realistic appraisal of the integrity of the containment under accident conditions "as found" local leakage results must be 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.
Amendment No. 27, M,63 4-34a
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