ML20138B695

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Proposed Tech Specs Replacing Pages 4-34,34a & 34b & Adding 34c to Reflect Exemption from 10CFR50.12,App J,Section III.D.3 Schedular Requirements for Local Leak Rate Testing
ML20138B695
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 12/09/1985
From:
GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20138B676 List:
References
NUDOCS 8512120361
Download: ML20138B695 (9)
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Text

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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 immedia tely. ,
b. If conformnce 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 effected and the local leakage meets the acceptance criterion as demonstrated by retest.

4.4.1.2.5 Test Frequency y Local leak detection tests shall be performed at a frequency as required by 10 CFR 50 Appendix J, except that:

a. The equipment hatch and fuel transfer tube seals shall be tested 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 mnths. When the airlocks are opened during the interim between six month tests, the airlock door resilient 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 series of openings. This requirement exists whenever containant integrity is required.
c. The reactor building purge isolation valves shall be leak tested each refueling interval per 10 CFR 50, Appendix J, Item III.D.2.
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 requirement is not in effect during cold shutdown.
e. Readings of the rotameters in each unifold of the penetration pressurization system shall be recorded at periodic intervals not to exceed three months.
f. Where an exemption from the frequency specified by 10 CFR 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.

G512120361 851209 '-

hDR ADOCK 0500 9 4-34 Amndent Nos, fd,108

t 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 my 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, riandestructive 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.

4.4.1.5 Reactor Building Modifications Any ajor 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 6 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 is inoperable, each entry and exit via that airlock shall be locally supervised by a menber of the unit operating mintenance 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 mintained during periods of interlock inoperability in an
appropriate station log.
3. If an interlock is inoperable for 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 Tech. Spec. 4.4.1.2.5d shall be performed 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 0.6LA. See Tech. Spec. 3.6.8 for further action.

4-34a Amendment Nos.f P/ , $, f)d,108

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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.

Bases (l)

The Reactor Building is designed for an internal pressure of 55 psig and a steam-air mixture temperature of 281F. Prior to initial operation, the containment was strength tested at 115 percant of design pressure and leak rate tested at the design pressure. The containment was also leak tested prior to initial operation at approxistely 50 percent of the design pressure. These tests established the acceptance criteria of 4.4.1.1.3.

The performnce 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 noral anner prior to local or integrated leakage rate tests.

The minimum test pressure of 27.5 psig for the periodic integrated leakage rate test is sufficiently high to provide an accurate measurement of the leakage rate and it duplicates the pre-operational leakage rate test at the reduced pressure. The Specification provides a relationship for relating the measured leakage 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 mjor considerations. First is the low probability of leaks in the liner, because of conforance 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  ;

(0.06 percent) of leakage that is specified as acceptable from penetrations and isolation valves. Third is the tendon stress surveillance program which provides assurance that an important part of the structural integrity of the containment is mintained.

4-34b Amendment Nos. $, (/3,108

l 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 0.06 percent from those penetrations 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 established 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 system is specified to assure continued operability and preclude instances where one or both doors are inadvertently lef t 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 performnce as containment isolation barriers.

References (1) FSAR, Section 5 4-34c

l GPU Nuclear Corporation Nuclear o

=eme:reo s Middletown, Pennsylvania 17o57-o191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Number:

October 22, 1985 5211-85-2167 n

Office of Nuclear Reactor Regulation Attn: Mr. H. Denton, Director

..r' r 0""X

'] ' /I 7 mi3 U.S. Nuclear Regulatory Commission Washington , D.C. 20555

Dear Mr. Denton:

Three Mile Island Nuclear Station Unit 1 (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 Exemption from Local Leak Rate Testing (LLRT)

Schedule Requirements As a result of delays in TMI-l's operating schedule, GPUN finds it necessary to request an exemption from licensing requirements in order to prevent a forced outage to perform tests that have been planned for the Eddy Current Outage which is now scheduled to begin in March,1986.

In accordance with 10 CFR 50, Appendix J, Section III.D.3, TMI-l is required to complete the first tests in the next series of Local Leak Rate Tests (LLRT) by February 23, 1986. Without prior approval of the NRC, TMI-l Technical Specifications (T.S.) also requires the first tests to be completed by that date. Barring an unforeseen shutdown of sufficient duration, these schedule requirements cannot be met without bringing the plant to cold shutdown conditions solely for the purpose of completing LLRT by their required dates.

Forced shutdowns during the Restart test sequence may require the plant to be in cold shutdown for short periods of time during which GPUN will attempt to complete as much of LLRT as possible. However, as an alternative to cooling down specifically to perform these tests, GPUN is requesting a one time exemption in accordance with 10 CFR 50.12 from Appendix J,Section III.D.3.

This exemption would allow completion of those tests which we are unable to perform by their required date to be delayed until startup from the Eddy Current Outage or from any forced cold shutdown prior to the Eddy Current Outage which is recognized to be of sufficient duration. In any event, if the balance of LLRT is not completed before August 23, 1986, GPUN would shutdown in order to complete them.

This one time only exemption pertains to LLRT, Type C Tests required by 10 CFR 50, Appendix J which are intended to measure the leakage rates of containment isolation valves (CIVs). Appendix J,Section III.D.3 requires that Type C tests be performed during each reactor shutdown for refueling but GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation

5211,-85-2167 October 22, 1985

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in no case at intervals greater than 2 years. T.S. 4.4.1.2.5 requires these tests to be performed at a frequency of at least each refueling period.

T.S.1.2.8 defines a Refueling interval as "the time between normal refuelings of the reactor, not to exceed 24 months without prior approval of the NRC."

It should be noted that T.S. allows for approval of this exemption without further amendment.

! Type C tests for which exemption is being requested are those tests which must

, be performed during cold shutdown conditions. These are tests which would require test equipment to be connected inside the reactor building, affect the ability of the system to support reactor operation, or tests that would take equipment out of service that is required by T.S. for reactor operation (that

is, action that would begin a T.S. time clock). Valves whose testing would fall into one or more of these three categories are listed in Table 1.

Valves CA-Y4A, CA-V13, and CM-V4 are not listed in Table 1 since these valves have recently undergone post-maintenance retesting. Valves IA-V6, IA-Y20, LR-V4, LR-VS, LR-V6, SA-V2 and SA-V3 can be tested during plant operation and exemption is not needed for tests of these valves.

Those valves listed in Table 1 and indicated by an asterisk are valves which have been added as part of a plant modification and do not appear in i

T.S. 4.4.1. 2.1.b. Under license condition 2.c, TMI-l's operating license is subject to all applicable rules, regulations, and orders at the time of issuance or thereafter in effect. Therefore, Appendix J test requirements may apply to valves not specifically listed in T.S. Currently the NRC Technical Specification Improvement Project (TSIP) and the AIF Subcommittee on Technical Specification Improvements are examining the appropriateness of duplication within the technical specifications of regulation such as Appendix J and I

whether or not technical specifications should include detailed component lists such as the component lists included in T.S. 4.4.1.2.1. Therefore, it I

is appropriate to await the resolution of these issues prior to amending l T.S. 4.4.1.2.1.b in order to update the valve list.

On May 29,1985, the NRC Commissioners voted to lift the shutdown ordcr which had remained in effect for over 6 years. In anticipation of NRC's authorization to restart, which GPUN felt was imminent, a plant heatup was conducted on June 7,1985 and the plant remained in hot shutdown. As the restart authorization from NRC. was delayed through the courts, our schedule has been pushed back in increments practically on a daily basis. During that time it became apparent that LLRT which had been scheduled to be done during the Eddy Current Outage could not be completed within the time frame allowed without performing a cooldown.specifically for that purpose and discussions were held with the NRC staff concerning the exemption which we are now requesting.

The results of the last LLRT, which were submitted to the NRC on July 19, 1984, show that the "as left" leakage was less than one third the total leakage allowed by T.S. Although TMI-1 has not operated since March 1979 order to maintain an operational readiness for Restart during that time we, in I

i

t 5211-85-2167 October 22, 1985 l

have completed four series of LLRT, which is more than would ordinarily be required over a 6-1/2 year span. From these and prior test results, GPUN has developed a historical data base from which we feel that those leakage paths which could become a problem have been identified and the necessary repair

! work has been accomplished with satisfactory results.

TMI-1 is presently engaged in Restart Test Program activities following the shutdown order which remained in effect for 6-1/2 years. The Restart Test Program includes zero and low power physics tests, natural circulation tests, and other tests in a slow controlled power escalation test sequence which is followed by about 90 days of operations prior to shutting down for the eddy current outage in compliance with License Amendment No.103. We feel that to place the plant in cold shutdown only for LLRT purposes would interrupt the Restart test sequence and 0TSG test run unnecessarily. It would also subject the plant to an unnecessary thermal cycle whereas we have a planned outage for eddy current testing which will occur within a reasonably short time after the original required date. Therefore, for the reasons stated above, we conclude that a limited delay in the schedule frequency requirements of the LLRT on a one time only basis as would be allowed by this exemption is justified and that in the interim, containment integrity can be assured.

In accordance with the provisions of 10 CFR 170.21, a check for $150.00 is attached as payment of the fee associated with the review of this request.

Sincerely, i

Director, TMI-1 HDH/MRK/spb:0377A

! Attachment

cc: R. J. Conte, Senior Resident Inspector, NRC J. Thoma, Operating Reactors Branch No. 4 T. E. Murley, Region I, Regional Administrator

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TABLE 1 Containment Isolation Yalves - Exemption from Appendix J Type C Test Schedule Depending upon the plant operating schedule, the valves listed below may require exemption from Appendix J test schedule requirements in order that they be tested during the planned outage:

Valve No. Size (in.) Last Dat'; Tested

1. CA-V1 1 03/15/84
2. CA-V2 1 03/15/84
3. CA-V3 1 03/15/84
4. CA-Y4B 1 02/27/84
5. CA-VSA 1 02/29/84
6. CA-V5B 1 03/01/84
7. CA-V189 2 03/26/84
8. CA-V192 2 03/26/84
9. CF-V2A 1 03/16/84
10. CF-V2B 1 03/15/84
11. CF-V12A 1 03/16/84
12. CF-V12B 1 03/15/84
13. CF-Vl9A 1 03/17/84
14. CF-V19B 1 03/17/84
15. CF-V20A 1 03/16/84
16. CF-V20B 1 03/15/84
17. CM-V1 1 02/23/84
18. CM-V2 1 02/23/84
19. CM-V3 1 02/23/84
20. DH-V64 2 03/04/84
21. DH-V69 2 03/02/84
22. HM-VlA* 0.5 03/20/84
23. HM-V1B* 0.5 03/18/84
24. HM-V2A* 0.5 03/20/84
25. HM-V2B* 0.5 03/18/84
26. HM-V3A* 0.5 03/20/84
27. HM-V3B* 0.5 03/18/84
28. HM-V4A* 0.5 03/20/84
29. HM-V48* 0.5 03/18/84
30. HP-V1 6 03/23/84
31. HP-V6 6 03/23/84
32. HR-Y2A 2 03/19/84
33. HR-V2B 2 -

03/19/84

34. HR-Y4A 2 03/19/84
35. HR-V4B 2 03/19/84
36. HR-V22A 2 03/19/84
37. HR-Y22B 2 03/19/84
38. HR-V23A 2 03/18/84
39. HR-V23B 2 03/18/84 l

l f

l

,o . .? E Valve No. Size (in.) Last Date Tested

40. IC-V2 6 03/09/84
41. IC-V3 6 05/13/84
42. IC-V4 6 03/10/84
43. IC-V6 3 03/11/84
44. IC-V16 4 03/11/84
45. IC-V18 6 03/10/84
46. LR-V1 6 03/23/84
47. LR-V10 6 03/23/84
48. LR-V49 6 03/23/84
49. MU-V2A 2.5 04/07/84
50. MU-V2B 2.5 04/07/84
51. MU-V3 2.5 04/07/84
52. MU-V18 2.5 03/29/84
53. MU-V20 4 03/13/84
54. MU-V25 4 03/14/84
55. MU-V26 6 03/14/84
56. MU-V116 1.5 03/13/84
57. NI-V26* 1 03/25/84
58. NI-V27 1 03/25/84
59. NS-V4 1.5 03/13/84
60. NS-V11 8 03/12/84
61. NS-V15 8 03/12/84
62. NS-V35 8 03/13/84
63. RB-V2A 8 03/27/84
64. RB-V7 8 03/22/84
65. SF-V23 8 03/25/84
66. WDG-V3 2 03/11/84
67. WDG-V4 2 03/11/84
68. WDL-V303 4 03/17/84
69. WDL-Y304 4 03/17/84
70. WDL-V534 8 03/24/84
71. WDL-V535 8 03/24/84
  • Valves which are required to be tested in accordance with 10 CFR 50 Appendix J but are not listed in T.S. 4. 4.1. 2.1. 6.

0377A

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