ML19347F084
| ML19347F084 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/11/1981 |
| From: | Hovey G METROPOLITAN EDISON CO. |
| To: | Barrett L Office of Nuclear Reactor Regulation |
| References | |
| LL2-81-0094, LL2-81-94, NUDOCS 8105150228 | |
| Download: ML19347F084 (3) | |
Text
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6-Metropolitan Edison Company
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gg Post Office Box 480 Middletown, Pennsylvania 17057 717 944-4041 May 11,1981 LL2-81-0094 "E
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TMI Program Of fice ig Attn:
Mr. Lake Barrett, Deputy Director N
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Middletown, Pennsylvania 17057
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Dear Sir:
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Three Mile Island Nuclear Station, Unit 2 (TMI-2)
Operation License No. DPR-73 Docket No. 50-320 Request for an Exemption from the Testing Requirements of 10CFR 50 Appendix J Paragraph (o) of 10 CFR 50.54 " Conditions of Licenses" states that
" primary reactor containment for water cooled power reactors shall be subject to the requirements set forth in Appendix J."
Appendix J specifies leak test requirements for verifying the leak-tight integrity of tN primary reactor containment. However, performing Appendix J tests at Three Mile Island Nuclear Station Unit 2 (TMI-2) is no longer appropriate with the reactor and the containment in their current condition. Therefore an exemption from the requirements of Appendix J is requested.
The basis of this request is the fact that potential containment pressurization modes would only result in very slight incremental positive pressure changes, commencing from a nega ive pressure as requiref7 paragraph 3.6.1.4 of the Recovery Technical Specifi-
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cations. For example, an analysis was performed to bound containment building pressure change in the event of failure of all the Reactor Building Air Coolers which are located inside the containment building. This analysis concluded that the pressure inside the containment building would take several days to increase by one to two psi, assuming this scenario occured during the summer months.
This analysis was based on the following heat inputs:
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a)
Solar Radiation = 1.82 X 106 BTU /Hr.
b)
Cota Decay Heat = 0.327 X 106 BTU /Hr. (This value has 6 BTU /Hr.as of 9D since decayed to approximately 0.11 X 10 May 1, 1981.)
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Conduction and convection through concrete walls (based on 900F and 800F outside and inside air temperature respectively) = 0.142 I 106 BTU /Hr.
Another analysis which was performed assumed instantaneous release of all reactor coolant to the containment.
In this scenario:
I a)
The average (bulk) incore temperature was assumed to be about 1200F, b)
The reactor coolant system fluid was assumed to be instantaneously and hoogeneously released, c)
The heat contained in the fluid was assumed to homogeneously distributec to all two million cubic feet of air in the containment d)
No credit for heat transfer to any components or the concrete structure was assumed.
The analysis yielded an approximate two (2) psi pressure increase
.n the reacter containment building.
The sole event which could cause containment pressure to exceed there low pressures is a recriticality accident. The probability of this accident was evaluated in the Programmatic Environmental Impact Statement which stated that "(t)he most probable (although very unlikely) cause of a recriticality 4
was found to be boron dilution, which would be a slow enough proces s that any approach to criticality can be detected and remedied."
Therefore based on the above, we believe there is adequate justification for granting an exemption from Appendix J testing. Additionally, performance of this testing has ALARA implications in that performance of the required testing would involve a considerable amount of work in high radiation areas.
Hence it would result in significant radiation exposure to test personnel.
Specific reasons for exemption from each type of Appendix J testing are discussed below.
Type A testing is intended to measure primary reactor containment overall integrated leakage under design basis accident conditions. However, as discussed above, because of the current conditions at TM1-2 the containment is subject to very low positive pressures in the event of an accident. Further, performance of Type A testing requires extensive preparation inside the containment prior te pressurizing the containment. With many of the areas of containment physically and/or radiologically inaccessible these preparations cannot be completed.
Additionally Nhc approved modifications have been made to containment penetra-tions R401, RS61 and R626, to allow their use during the recovery, which have reduced their design pressure. For these reasons it is currently neither necessary, nor possible, to perform a Type A test.
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,.. o Mr. Lake Barrett Type B testing is intended to detect local leaks and to measure leakage across each ' pressure containing or leakage limiting boundary for reactor containment penetration, such as the containment air locks and resilient seals. There
' presently is no pu pose in testing these penetratio s to 56.2 psig (Pa) as required by Appendix J because, as discussed previcasly, the containment is subject to only very low positive pressures in the. vent of an accident.
Additionally, although the penetration pressurization connections for all containment penetrations are accessible, approximately half of the penetrations are located in high radiation areas outside the containment. Therefore even if a leak.is detected from any of these penetrations there is a hign probability that the source of the leakage could not be found and repaired. Further, the unmodi-fled penetrations form a double barrier with a design pressure of 60 psig which is more than adequate for the very low positive pressures to which the contain-ment may be subject. Therefore type B testing need not be performed.
Type C tests are intended to measure containment isolation valve leakage rates.
In this test the containment isolation valves are tested for leakage against a test pressure (Pa) of 56.2 psig is much greater than the very low pressures the containment is subject to in the event of an accident. Further the Recovery Technical specifications require us to maintain two OPERABLE containment isola-tion valves closed when not required open in accordance with an approved procedure. This forms a double barrier with a design pressure of 60 psig (for these isolation valves) which is significantly greater than the maximum possible containment pressure as discussed earlier. Therefore type C testing need not be pe rfo rmed. Additionally 64 of 67 mechanical penetrations require operations to be performed in high radiation areas both inside and outside of containment which would result in sfgnificant personnel exposures.
In conclusion: Type A testing is neither necessary, nor possible, under current conditions and Type B & C testing would serve no useful purpose as the containment is at most subject to very low positive pressures which are much lower than the design ' pressure of the subject items. Additionally,there is a very high prob-ability that any leak that is discovered will be in a high radiation area which is not accessible and disassembly of components required to fix the leak could result in an unsafe condition. Therefore we request that an Appendix J exemption be granted. In the event a decision is made to restore TMI-2 to an operable condition we understand that we will then be required to comply with the require-ments set forth in Appendix J.
Sincerely,
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G. K. Hovey Vice President and Director, TMI-2 GKH:JJB:be cc:
Dr. B.J. Snyder, Program Director - TM1 Program Of fice
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