ML19292A991

Provides Clarification of Board Notification 83-47 Re PORVs for Transmittal to Licensing Boards Associated W/Plants Designed by Westinghouse & B&W

ML19292A991
Person / Time
Site:	05000000, Indian Point
Issue date:	06/22/1983
From:	Mattson R Office of Nuclear Reactor Regulation
To:	Eisenhut D Office of Nuclear Reactor Regulation
Shared Package
ML19292A992	List: ML19292A991
References
NUDOCS 8307130133
Download: ML19292A991 (1)

Text

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BSheron LMarsh Jim 2 21993 MEMORANDUM FOR: Darrell Eisenhut, Director Division of Licensing FROM:

Roger Mattson, Director Division of Systems Integration

SUBJECT:

BOARD NOTIFICATION ON PORVs

Reference:

Memorandum from R. Mattson to D. Eisenhut, " Board Notification Regarding PORVs," thrch 27, 1983.

In the reference memorandum, we requested you notify licensing boards associated with plants designed by Westinghouse and Babcock and Wilcox of our recent determination regarding the need for PORVs to be classi-fied as safety-related equipment.

Since the issuance of the reference memorandum and your subsequent board notification, we have determined there is a need to clarify our position on this issue.

We have recently issued a memorandum to the Chairman of the CRGR in response to his inquiry as to whether or not our origina. Soard noti-fication position constituted a new generic requirement. Our response to this question is provided as enclosure (1) and provides the necessary clarification of our position. The resul y of the CRGR review of this clarification has recently been issued and is provided in enclosure (2).

I request you provide enclosures (1) and (2) to the affected licensing boards so that they are aware of our clarification and the CRGR review of our clarification of this issue.

Original s: ped by Roger J. Mattian Roger J. Mattson, Director hg Division of Systems Integration v

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MEMORANDUM FOR: Darrell Eisenhut Director, Division of Licensing FROM:

Roger Mattson, Director, Division of Systems Integration

SUBJECT:

BOARD NOTIFICATION ON PORVs

Reference:

Memorandum from R. Mattson to D. Eisenhut, " Board Notification Regarding PORVs," March 27,-1983.

In the reference memorandum, we requested you notify licensing boards associated with plants designed by Westinghouse and Babcock and Wilcox of our recent determination regarding the need for PORVs to be classi-fied as safety-related equipment.

Since the issuance of the reference memorandum and your subsequent board notification, we have determined there is a need to clarify our position on this issue.

We have recently issued a memorandum to the Chairman of the CRGR in response to his inquiry as to whether or not our original board noti-fication position constituted a new generic requirement. Our response to this question is enclosed and provides the necessary ularification of our position.

I request you provide the enclosed memorandum to the affected licensing boards so that they are aware of our clarification of this issue.

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Roger J. Mattson, Director Division of Systems Integration

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Enclosure:

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Background===

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Over the past few years, the. staff _has been studying the systems-related

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SRP Guidance Section 15.6.3 of the Standard Review Plan provides the guidance for reviewing the " Radiological Consequences of Steam Generator Tube Failure".

As stated in the title, the guidance provided primarily covers the review of the radiological consequences and not much emphasis is placed on the details of the systems review.

Section III.2 of the review procedures states that "The AEB reviewer verifies with the RSB the acceptability of the applicant's description of events, including operator actions, to assure that the most severe case has been considered with respect to the release of fission products and calculated doses".

The RSB review usually did not address assumptions made regarding equipment operability associated with overall system thermal-hydraulic behavior.

2_

Credit for Non-Safety Related Ecuipment The staff has historically had a policy which, in general, required equipment relied upon to mitigate the consequences of design basis accidents to meet sa fety-related requirements.

See, for example, the acceptance criteria listed for " Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break", SRP Section 15.3.4 Exceptions have been made on a case by case basis however.

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position) thiat the component or system in questi$n would. perform its intended function in a reliable manner, despite its non-safety related classification.

On November 20, 1981, the Director of HRR issued a memorandum to all NRR personnel setting forth " standard definitions for ccmmonly-used safety classification terms."

In this memorandum, it was stated that "...The definitions.... should be considered ' standard' and should be applied consistently by all NRR personnel in all aspects of our safety review and, licensing activities...." Under the definition of " Safety-Related,"

item 3 (from 10CFR100, Appendix A) states a safety-related structure, system or component is one which has "...the capability to prevent or mitigate the consequences of accidents which could resi.?t in potential

off-site exposures comparable to the guideline exposures of this part (Part 100)."

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.).2 Q'f;";...R: event;to' depres~surize the priinary s'ystem.pfN ...'SGT ~ i secondary leak }within;.30 minutes.Je do not<know whether" credit 36E PORVN .:;p& . 4%G5:: =%q, 3,ggig..g";g: ,,_,,;,.e;.gy.! ?.jQp a ; 4.;.9 ~ M-x ._.g,. y.g... ..y n,...., ' operability is necessary in order'to demonstrate, su'c'cessful mitigadon Y ' '. $'T of the SGTR event. '. --. s s., We propose the following: 1) Each applicant would be requested to determine if PORV operability is required in order to successfully mitigate the SGTR event (i.e., demonstrate offsite doses are within 10CFR100 guideline values). 2) If PORV operability is required, then in accordance with the 11/20/81 guidance memorandum on safety-related equipment, the PORV ,would be classified as a " safety-related" component. The staff would then conduct its review to determine if the design criteria for safety-related equipment are met by the proposed PORV(s). If all of the design criteria associated with safety-related equipment

_4 were not met, then the applicant would be requested to provide justification why credit should be given for.the PORY to perform its safety function in a reliable manner under the accident conditions in question. If suitable justification is provided, the staff would lg accept it on a case by case basis. = '.,y _ ,.w., ' W The implementatio&Q. '.:ll.]ll,: (.;.y @'dered either,a. ...' w: M T fik.& & yWfy.f.f r'l %pgop# ;.A M %jis not.consii N A 3 rig Q $ h % fbb 4,?y.

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4. generic requirement.pThe staff hasihistorically, required.that equipment v

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1: - w . p t t, -; ...\\Q' W=~ - ;.19S ' a::-.4.Yd55%..li2E@.l"..?..Y '. 59l;:,j. requir 7.- u: 0 ~ s?.:.'& .. 7 3 ~. ' recognized the significance of.the systems aspects of. SGTRs until'.1

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- ? -..m- .w.Qc. . p.( y, -y recently. In NRR Office Letter.No. 39 (Revision 1) it was, pointed out that CRGR review was not necessary for case-specific applications of existing requirements. I have determined that this is the case for this item. If an applicant wishes to take credit for equipment necessary to mitigate postulated accidents that is not fully designed to safety-related criteria, then the applicant should submit justification why it is acceptable to do so. The staff will review this justification, and, if found acceptable, will allow appropriate credit in its safety evaluation. This approach is consistent with previous staff practice and would be conducted on a case-by-case basis.

5 Additionally, the steam generator generic requirements, currently undergoing final NRR review, contain a specific task for the staff to re-examine and revise the steam generator tube rupture (SGTR) standard review plan (15.6.3). This task.will involve an assessment of the SRP analysis, single failure and.,offsite power requirements, as well as.a reexamination of the overall manner in which this accident is reviewed within the NRC.. It is. currently envisioned that more rigorous RCS. .... w - w* e

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-,n .S OLs ..y.". It is our understanding that Westinghouse is now recommending electric + solenoid operators for the PORVs rather than pneumatic operators. The following 01. plants have electric solenoid operators for their PORVs. These operators will be safety-related and environmentally qualified. Seabrook 1/2 Beaver Valley 2 Vogtle 1/2 Millstone 3 Callaway Wolf Creek Diablo Canyon 1/2

The following Westinghouse plants have not yet received a, license and have pneumatic operators for the PORV which we understand currently do not meet safety-related design criteria and are not environmentally qualified. Comanche Peak 1/2 k - ^ . J., Byron 1 .c. c. 9.- . y$ttg Bav 1/2 ai W,,.A.. o ':g - .- p ,v ^ . j. l~ C :.. >,., .. Qh. l- ' * ' }h@.'M*i. Y.h. Q. Q, ',j,E:'*. ".Qil;.1f&. i.ji%.,.pi.e.vM (.y,.,h.f..,.,l, f.; k '. x..'.U %..'-l. E$ Q$'). 5?i'. ' ' ~ h,, :,. [:',* 7 .,.\\ i... 9,.r. = .. :y. tawba.1 ':... Ca . ;. -::4Vry.f'A. c~m-: s. - -m.w m2 * . q~* '. f%.. -.,.,;&

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~ ~ ~ ' ~. There are currently 3 B&W plants with applications pendihg for an OL. These plants are Midland, Bellefonte, and WNP-1. We are aware that Midland has a safety-related PORV and operator according to their FSAR and thus is not affected by board notification BN-83-47. We have not yet determined the extent to which the PORVs in Bellefonte and WNP-1 can be considered safety-related. For example, we are not sure if WNP-1 can meet the single failure criteria for PORV operability. All CE-designed plants presently applying for OLs' rely on safety-related auxiliary pressurizer spray to accomplish depressurization and currently do not have PORVs in their design. Therefore the issue is not applicable to these plants.

OR's We have not performed a survey to deterTnine how many OR's do not have safety-related PORVs. However, we propose to address this issue in the same manner as other potential backfit requirements. We will identify it as a potential generic issue. The Safety Program Evaluation Branch in DST will be requested t,o prioritize it and resources will be assigned ~ to this issue for resolution'comensurate with the priority assigned;,. ... r....y. ;..r. : n.,. q.. ;,,.. _ v ; -,.... ~.. .w .c. .,. :_, e:.7.:T.d'n;en;,u.rj?pl:dl4Ci2.b%. r c.e.p. naqqyrm.a-i.g. i5.;54i%'a.9 ;i;).2.Th,k.f n. . v. . 7 s. c ..., : x.

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UNITED STATES g y} g NUCLEAR REGULATORY COMMISSION

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fiEMORANDUM FOR: Harold R. Denton, Director Office of Nuclear Reactor Regulation FROM: Victor Stello, Jr. Deputy Executive Director Regional 'Operati.ons and Generic Requirements

SUBJECT:

CRGR REVIEW OF PROPOSED REQUIREMENT FOR PRIMARY SYSTEM DEPRESSURIZATION CAPABILITY I have reviewed your memorandum of May 31, 1983, on this subject. In that memorandum (pages 3 and 4) you redefined the proposed requirement presented in Board Notification BN-83-47. The proposed requirement (as redefined) would require each applicant "to determine if PORV operability is required in order to successfully mitigate the SGTR event (i.e., demonstrate offsite dosesarewithin10CFR100guidelinevalues)." Following that, if PORV operability is required, applicants would be required to upgrade the PORV or provide a justification. The first part of the proposed requirement which calls for an assessment of PORY operaaility requirements is not considered to be a new requirement or a new interpretation of an existing requirement, and is therefore exempt from CRGR review as provided for in Section III.B.i of the CRGR Charter. The second part of your proposal includes a requirement for an applicant to provide a justification for assuming performance of a nonsafety grade PORV, which the staff would then approve or disapprove on a case-by-case basis. We believe that any resulting generic requirtment to upgrade equipment would constitute a new interpretation of an exic.!ng requirement. Therefore, a review would be required in accordance with Section III.B.i of the CRGR Charter. There is one other matter that I believe you should consider. Board Notification BN-83-47 issued April 4, 1983, is not consistent with your May 31, 1983 memorandum and probably should be revised. I understand from my staff that this concern has been conveyed to your staff and an activity is under way to clarify the Board Notification. / ,4-n4h m

c JUN 141933 I suggest that you keep Tom Cox of my staff informed of the status of this matter. He may be reached on extension 24357. s' ~ 1ctor Stello, Deputy Executive Director Regional Operations and Generic Requirements cc: W. Dircks G. Cunningham CRGR Members ( 9 4}}