ML20072T242
| ML20072T242 | |
| Person / Time | |
|---|---|
| Site: | Salem, Hope Creek  |
| Issue date: | 04/11/1991 |
| From: | Preston B Public Service Enterprise Group |
| To: | NRC OFFICE OF THE SECRETARY (SECY) |
| References | |
| FRN-56FR03796, FRN-56FR3796, FRN-57FR34666, RULE-PR-50 56FR3796-00002, 56FR3796-2, AD05-2-023, AD5-2-23, AD5-2-54, NLR-N91060, NUDOCS 9104180037 | |
| Download: ML20072T242 (3) | |
Text
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'O 000KET HlAGER
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Pub::c Service Electnc and Gas Company P O. Box 236 Hancocks Snoge. New Jerr,ev 06038 Nuclear Department 9I P^
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Secretary, US Nuclear Regulatory Commission Docketing and Service Branch Washington, DC 20555 Gentlemen:
COMMENTS ON PROPOSED CHANGE TO 10CTR50.55A INSERVICE TESTING OF CONTAINMENT ISOLATION VALVES SALEM AND HOPE CREEK GENERATING STATIONS DOCKET NOS. 50-272, 50-311, AND 50-354 Public Service ElJctric & Gas Company (PSE&G) hereby submits comments for the proposed rule change to 10CFR 5,0.55a. contains comments on paragraph b(2)vil - Inservice Testina of containment Toolation Valves and paragraph (6) (11 ) ( A) ( 3 ) Augmented Examination of Reactor Vessel.
Should there be any questions with regard to this submittal, please do not hesitate to contact us.
Sincer ly, f nm y
B. A. Preston Manager -
Licensing and Regulation l
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~Md I iM ATTACRMENT 1 NLR-N91060 COMMENTS TO 10CFR50.55A PROPOSED RULE CHANGE Paragraph b(2)(vii) - Inservice Testing of Containment Isolation Valves The proposed rule, through the adoption of the 1989 Edition of ASME Section XI, adopts ASME/OM-10 for valve testing.
This specific paragraph takes exception to certain requirements of OM-10.
Specifically, containment isolation valves must be analyzed in accordance with paragraph 4.2.2.3(e) and corrective action taken in accordance with paragraph 4.2.2.3(f) of ASME/OM-10.
However, the ASME/OM-10 standard refers all requirements for containment isolation valves to 10CFR50 Appendix J.
Containment isolation valve leak rate testing is required by 10CTR50 Appendix J.
These requirements, implemented through technical specifications, require valvo leakage to be evaluated against acceptance criteria.
Unacceptable leakage rates require valves to be repaired, leakage paths to be isolated, or operating units to be shutdown.
In addition, if unacceptable leakage rates are identified during outages, plants are prevented restart.
If the proposed rule is adopted as written, it would require utilities to consider Technical Specification actions, in addition to those identific:d in the preceding paragraph, which may may force plants to shut down.
If they evaluated leakage in accordance with Appendix J, they would not have to.
There have been reports published on excessive leakage through containment isolation valves and the effect on containment leakage integrity.
One example is a paper titled, " Primary containment Leakage Integrity: Availability and Review Experience"; published in Nuclear Safety, Vol. 21-5, dated September-October 1980.
This report identifies many of the valve leakage failures to be repetitive in nature with these problems associated with large diameter MSIV's, purge or ventilation valves.
A s a result cf reports such as these, the generic problems with these valves are being addressed and individual valve leakage acceptance criteria for these problem valves have been identified in plant Technical Specifications.
The remaining problems with valve leakage are more than likely plant specific and are identified during Appendix J, Type C Testing.
The Appendix J program has proven to be effective in identifying containment isolation valvo leak rate problems.
t.-
41 mbt NLR-H91060 It is recommended that any revisions the NRC fools is required to containment isolation valve leak rate testing requirements be made in conjunction with Appendix J.
By duplicating valvo leak rate acceptance criteria in the plants IST Program, plant operability will be challenged more often without a corresponding increase in safety and it will add confusion in Technical Specification implementation.
Paragraph (6) (ii) ( A) (3) - Augmented Inspection of Reactor Vossol This paragraph allows plants with fewer than 40 mor,ths reliaining in the Inservice Inspection interval to defor the augment <ad examination of the Reactor Vessel shell welds to the fir 9t period of the next interval.
The te"m augmented examinatinn refers to the examination of 100% of the shell wolds, whereas,Section XI requires only a partial examination of theco welds.
This paragraph also prohibits the uso of the deferred augmented examination as a substitute for the reactor vessel shell examination scheduled for implementation during the current inspection interval.
Prohibiting the use of the deferred augmented examinations as a substitute for the reactor shell examinations scheduled within the next 12 months would causo additional expenditure of radiation exposure, time, and money within a short period of time.
It would be too late to adequately plan to perform the augmented examinations during the next scheduled outage, even though the reactor vessel core barrel would be removed to complete the examination requirements of the current interval.
Since the coro barrel must be removed to perform the augmented examinations required by this rule chango, PSE&G would thoroafter be forced to remove the core barrel again within the subsequent 40 month period.
It has always been recognized within the code that removing the core barrel is a massive undertaking and performed only onco in 10 years.
It requires removing all fuel from the vessel, lifting the core barrel partly out of the reactor cavity water resulting in high exposure levels on the refueling floor, adds days to the duration of the outage, and significantly increases the cost of the outage.
Based on the above information, we recommend the proposed rule be revised to allow plants within the last 12 months of their current interval be allowed to substitute the deferred augmented reactor vessel shell examinations for the remaining reactor vessel shell examinations scheduled in the current interval.
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