ML19208B580
| ML19208B580 | |
| Person / Time | |
|---|---|
| Issue date: | 08/24/1979 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Kennedy R NRC COMMISSION (OCM) |
| References | |
| NUDOCS 7909210029 | |
| Download: ML19208B580 (16) | |
Text
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UNITED STATES y ;' "y, g
NUCLEAR REGULATORY COMMISSION 5,
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AUG 2 41979 11EMORANDUM FOR:
Commissioner Kennedy THRU:
ee V. Gossick, Executive Director f fe@atMf#
k H. R. Denton, Director, Office of Nuclear Reactor FROM:
Regulation
SUBJECT:
UCS PETITION FOR RECONSIDERATION By memorandum dated August 7,1979, you requested a response to certain questions concerning the UCS petition for reconsideration.
Our response is enclosed.
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$< G. &:t H. R. Denton, Director
'v0ffice of Nuclear Reactor Regulation
Enclosure:
As stated cc:
Chairman Hendrie Commissioner Gilinsky Commissioner Bradford Commissioner Ahearne L. Bickwit, GC A. Kenneke, OPE J. Fouchard, OPA C. Kammerer, OCA S. J. Chilk, SECY Union of Concerned Scientists
Contact:
E. Butcher X-28077 LEl8..UA 79 09 21002c?,.
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RESPONSE TO C0mISSIONER KENNEDY'S QUESTIONS DATED AUGUST 8, 1979 CONCERNING
_TjiE UES PETITION FOR RECONSIDERATION I.
IEEE Standard No. 323 - 1971 Version In order to facilitate a response to questions I & II, a table comparing IEEE Std. 323-1971 and IEEE Std. 32?-1974 has been prepared (see Attachment).
This table illustrates that with a few notable exceptions-(i.e., aging, margins, and detailed requirements for maintaining documentation) both of these standards address all of the major aspects of equipment qualification.
The principal difference between the two is the level of detail provided in the 1974 version as opposed to the interpretations and judgements required in the implementation of the 1971 version.
However, even with its greater detail, IEEE Std. 323-1974 still requires a significant amount of engineering judgement in its implementation especially in the area of aging and margins.
The significant point in these introductory remarks is that neither the 1971 version nor the 1974 version of IEEE Std. 323 can be applied effective-ly without guidelines for the necessary interpretations and judgements.
It is important to note that in the current reviews of license applications referencing IEEE Std. 323-1971 the staff is requiring interpretations and making judgements that bring the level of assurance of equipment qualification in these plants to essentially the same level as that which will be achieved in future plants from an application of IEEE Std. 323-1974.
O ih!. G.3
t COMPARIS0N OF IEEE STD 323-1974 AND 323-1971
,1974 1971 1.
Equipment Specification Specific requirements General requirements includiag:
a) perfcmance characteristics b) voltage, frequency ranges c) installation d) maintenance e) design life f) control, indication g) environment h) operating cycle i) qualified life 2.
Principles of Qualification Specific requirements General for regarding:
type tests, partial type a) type tests tests and analyses and b) partial type tests operating experience and analyses c) operating experience d) equipment interface 3.
Documentation Specific and detailed Less specific and requirements for diff-detailed with no erent qualification specific requirement methods.
Requirement to maintain documenta-to maintain documenta-tion files tion files.
4.
Type test Qualification Specific test plan, General test plan in Procedure including:
the form of data require-a) mounting ments for:
b) connections a) mounting c) monitoring b) connections d) margin c) monitoring e) test sequence d) test sequence f) aging e) vibration g) vibration f) radiation h) reiiation g) operation i) operation h) acceptance criteria j) inspection NOT INCLUDED AW::
k) acceptance criteria a) aging b) margin c) inspection
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1974 1971 5. Operating Experience Qualifi-Specific Ceneral outline of cation Method procedure in the form of data tequirements. 6. Qualification by Analyses Specific General outline of Procedure procedure in the form of data requirements. 7. On-Going Qualification Specific Not included i Procedure 8. Simulated Service Condition Description ard Not included Typical Test Prof.ile figures showing: a) margin (additional peak temp., press. and time) b) operation c) pressure and temperature d) time period \\ 1 4 !t ) " W . gb
2-These interpretations ar.d judgements are part of the continuing process of evolution of equipment qualificatica requirements discussed in NUREG 0413, Appendix A, " Report on the Historical Evolution of Environmental Qualification Requirements for Safety-Related Electrical Equipment." Question a How many operating nuclear power plants are not now formally committed to comply with the provisions of this Standard? '(
Response
Fiftyeight of thc 70 power reactors currently licensed to operate (including Indian Point 1 and Humboldt Bay) have no specific reference to IEEE Std. 323-1971 as the basis for equipment quelification. The licensees for the remaining plants have commitments to c' ply with the standard. Question b What are the adiantages to be gained by requiring those facilities - to formally cor.p,1y with the provisions of this Standard? the disadvantages? Rg ponse , In o'rder to provide a meaningful response to this question it must be answered within the context of the current ongoing staff program to upgrade \\ the qualification of electrical equipment in all operating reactors. This s program was outlined for the Commission at a briefing on July 11, 1379 to d'.scuss the licensee resodnses to IE Bulletin 79-01. The end result of this program will be a higher level of confidence in equipment qualifica-tion at operating reactors. This higher level of confidence will be cchieved by evaluating existing qualification inform (tion and documentaticn _ br corpliance with staff guidehnes to be establishea These guidelines 4 h h g L will provide a level of confidence essentially equivalent to that which would be achieved from the application of IEEE Std. 323-1974 whi h is more stringent than the earlier version of the standard. c Tnerefore, there is little advantage to be gained from limiting compliance to IEEE Std. 323-1971. The current staff program is intended to provide a higher level of confidence in equipment quali-fic? tion than one would get from the standard alone without also specifying guidelines for implementation. In fact there would be a disadvantage in that IEEE 323-1971 does not require consideration of eging or margins, nor would it include any specific requirement to maintain qualification documentation; all of which will be included in the staff's guidelines. _uestion c Q Based on the response to I.B. above, if such compliance were required by the Commission, wnat would be a reasonable schedule for demonstration of such compliance?
Response
At the Commission briefing on July 11, 1979, referred to in response to Question a. above, the staff stated that the reviews of licensees responses to IE Bulletin 79-01 against staff guidelines being prepared, could be completed by March 19A. iubsequent work in connection with preparing the guidelines %. " the scheduled da.te may be optomistic. Further evaluation of t' scheu e will be possible in October when the staff guidelines are available and thc interim evaluation of licensee responses is complete. Sirce the reauirements of IEEE Std. 323-1971 are less comprehensive than the staff': guidelines, a reeva'uation using this standard could be accomplished within the same time frame. O iliib9
II. IEEE Standard No. 323 - 1974 Version Question a What are the advantages to be gained by requiring operating nuclear power plants to comply with this Standard? the disadvantages?
Response
As noted above, the staff program for reassessing the adequacy of equipment qualification at operating plants will be based on guidelines which closely follow the requirements of IEEE 323-1974. Therefore, equipment that is found to be qualified in accordance with these guidelines should comply with most aspects of that Standard. Nevertheless, there are advantages to requiring that operating plants comply fully with all requirements of the 1974 standard. Specifically, the application of this standard to operating plants would require a more rigorous and complete demonstration that aging effects were acGuately accounted for and the establishment of a qualified life of all safety related equipment. Further, margins would be applied to all test parameters whereas the guidelines would require only that margins be applied to the most significant parareters (e.g., time). Implementation of both the margin and aging requirements would result in a somewhat hiaher level of assurance of the adequacy of qualification. The significant disadvantage of requiring licensees with operating plants to comply with IEEE 323-1974 is that many licensees will be required to retest some fraction of their equip ent to comply with the aging requirement. Some retesting may also be required to comply with more a conservative applica-tion of the margin requirenents. Where testing is not required, ONjOO compliance may also be demonstrated by analysis. In either case, significant licensee and staff resources would be required to implement these provisions and review the documentation. In addition, backfitting the standard would extend the date by which the ongoing reevaluation program could be completed and possibly delay the overall upgrading of equipment qualification that the staff program is designed to achieve. Question b Based on the response to II.a. above, which specific provisions of this Standard strike a favorable value-impact balance with respect to backfitting considerations?
Response
We believe that neither the aging nor the margin requirements of IEEE 323-1974, when compared to the draft guidelines to be used in the upcoming staff reassessment of operating plants, warrant backfit con. sideration. The benefit of backfitting either the aging or the margin requirements of the 1974 Standard is a small, unquen-tifiable increase in the level of assurance that equipment is qualified. Yet the costs, in terms of manpower, the testing required to implement these provisions, and the possible delay in the staff review effort may be significant. The staff guidelines mentioned above will require that aging be considered, but only for that equipment identified as being susceptible to significant aging effects. We believe that approach is adequate for the present. We are now reviewing operating appli-cations for plants already required to comply with the 1C74 version 3 /8.l7). of the standard and therefore we are assessing aging data for electrical equipment generally. As that review process proceeds, and our understanding of aging effects improve, we will reassess the need for further backfitting the aging requirements. With regard to margin, we believe that the application of the staff guidelines to operating plants will ensure that adequate margins have been applied during qualification testing. We believe that the more conservative application of margins required by the 1974 Standard, while appropriate for plants yet to be licensed, is unnecessary for operating plants. Question c Based on the response to II.b. above, what would be a reasonable schedule for demonstration of compliance with these provisions?
RESPONSE
We believe compliance with the 1974 version of the standard could be demonstrated in 3 to 4 years given there are adequate test facilities available during that time. sy ; q ) si x 40 s) 8 U.L 5 %*
III. Program for Independent Verification of Eauipment Qualification by NRC Question a What are the staff's recommendations for such a program?
RESPONSE
Sandia Laboratories, under an NRC contract, recently completed an analysis of alternatives for conducting independent verification testing for environmental qualification of electrical equipment. The staff is currently reviewing the Sandia report and we expect to submit our recommendations to the Commission by the end of September 1979. .sa
IV. Fire Protection Question a What are the staff's views regarding a rulemaking action to revise General Design Criterion No. 3 to explicitly establish, as acceptance criteria, confonnance with Branch Technical Position 9.5-1 and its Appendix A?
Response
The staff does not believe that this would be appropriate except for selected areas regarding fire protection as noted below. Experience with incorporation of detailed requirements and guidelines into regulations has shown that the advantage of specificity in requirements is counterbalanced by inflexibility in detailed application. Neither unforeseen plant-specific features nor improvements in knowledge can be incorporated into such rules in a timely way or without inordinate NRC resources. We therefore favor, in most technical areas, regulations setting forth the goal or principle, with implementation and detailed guidance embodied in Regulatory Guides or Standard Review Plans. However, we recognize the need for more specific rulemaking on some topics. The Office of Nuclear Reactor Regulation has requested the Office of Standard Development to proceed with a rule to explicitly state the minimum require-ments for the site fire brigade shift size and training requirements. We have concluded that the minimum fire brigade shift size at all operating reactor sites should consist of five trained members. Further, we have established a minimum acceptable level of fire brigade training in cases where licensees do not train all members in conformance to staff guidelines. The majority of licensees of operating reactors have committed to provide a trained $/8.1.N
- la - fiye-man fire brigade, however, there are a number who have not made such a comitment. OELD has advised that a rulemaking procedure is the most appropriate method to resolve this issue. The licensees are being advised that this procedure has been initiated. In the future we may determine a need for additional rulemaking similar to the current actions relative to fire brigade requirements at which time we will take similar actions. 3/bl.iji3
2 Question b What is the current status of the staff's reviews of fire protection programs at nuclear power plants?
Response
There are 70 plants licensed to operate. Since Humboldt Bay 3 and Indian Point 1, are currently not in service, safety evaluations are not now scheduled for these plants. The safety evaluations for 60 plants have been issued and the safety evaluations for the remairing 8 plants will be issued by Ssptember 15, 1979. Question c When is the earliest projected time that all operating nuclear power plants will be in full compliance with Branch Technical Position 9.5-1 and its Appendix A?
Response
Our initial date for the completion of plant modifications at operating plants was scheduled to be completed by October 1980. Most modifications will be implemented by that time; however, certain modifications cannnot be made because of the time required for the design and procurement of components and, in most cases, the timing of a refueling shutdown which is needed to complete the installation. These modifications will be completed during the plant's first refueling outage in 1981. Further, each safety evaluation contains some incomplete items which may result in additional modifications. Some modifications identified in the resolution of these incomplete items may not be done by October 1980. To expedite the completion of these modifications, we have recently {J/8.Ub
.. requested licensees to respond to the incomplete items within 90 days and we urged them to apply their best efforts to improve their schedules such that as many modifications as possible can be completed prior to October 1980. For several of the SEP plants, certain modifications, usually involving alternate or dedicated shutdown systems have been deferred until the com-pletion of the SEF evaluations so that the modifications can be used to resolve SEP issues also. Assuming that the SEP evaluations are completed early in 1981 and that some modifications to shutdown systems would be required, the completion date for installation of modifications for all operating reactors including SEP plants is 1983. Question d Are there any particular circumstances, eithe ceneric or plant-specific, which preclude earlier implementativ., ?ctions to achieve such compliance?
Response
There are four major reasons which preclude the earlier implementation of all modifications: 1. Dedicated or alternate shutdown systems could require as long as three years to design, construct and place in operation. 2. Some major modification require that the plant be shutdown when they are made and the first outage after the design and procurement is complete occurs after October 1980. 3. Equipment deliveries that are incompatible with the October 1980 date. 4. Sc ne major modifications have been deferred on the SEP plants. a ;M/7
Question e What is the present fire protection research program?
Response
The primary emphasis in the fire protection research program is currently being placed on the testing of full-scale mock-ups of four plant areas. Confirmatory testing will be conducted on detailed repli-cates of selected operating plant configurations, including the proposed fire protection designed in accordance with NRC guidelin2s and found acceptable by the staff. The purpose of these replication tests is to confim whether the fire protection measures proposed for the existing equipnent configurations are valid for operating plant conditions. Four plant areas have been chosen for replication testing. The areas, and the order in which they are to be tested are as follows: 1. Rancho Seco !iakeup Pump Room 2. Arkansas 1 Auxiliary Building Corridor 3. Brunswick Intake Structure Basement 4. Browns Ferry Reactor Building Considering the limitations imposed by available testing facilities, only a small portion of each of the above plant areas will be tested. The portion of each area to be tested will be chosen to confim the adequacy of the fire protection measures for the worst-case credible exposure fire to redundant safe shutdown system components in the area. M S.$.'l 0
The replication test working group, consisting of representatives of NRR, RES, Sandia and staff consultants, has met to scope out the test plans for the Rancho Seco makeup pump room test. Background information has been obtained from the licensee and distributed to the group. The working group will make a site visit to Rancho Seco in September to choose the specific portion of the area for testing and develop more detailed test plans. Additional elements of the fire protection research program are as follows: A. Fire Suporession Tests Suppression of cable fires with water, carbon dioxide and Halon to pmvide NRC with data on the effectiveness of these fire suppression agents on deep-seated fires. B. Cable Aging Tests Cable fire testing to detemine the effectiveness of fire retardant cables that have aged. C. Detection System Tests Tests of the methods used to position fire detectors in fires areas and tests of the sensitivity of fire detectors to the products of combustion of materials found in nuclear power plants. D. Separate Effects _ Tests Small-scale testing to verify the simulation methods used in the full-scale replication tests. E. Safety Related Eauipment Tests Tests of the response of safety related equipment (other than cable) when subjected to an exposure fire. O et.CO}}