ML20055J378

Summary of CRGR Meeting 168 on 890823.List of Attendees Encl

ML20055J378
Person / Time
Issue date:	09/06/1989
From:	Jordan E Committee To Review Generic Requirements
To:	Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
References
NUDOCS 9008020171
Download: ML20055J378 (85)

Text

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MEMORANDUM FOR:

James M. Taylor-Acting Executive Director for Operations September 6, 1989 O

FROM:

. Edward L. Jordan, Chairman Committee to Review Generic

[bb Requirements i

SUBJECT:

MINUTES OF CRGR MEETING NUMBER 168 The Committee to Review Generic Requirements (CRGR) met on Wednesday, August 23,1989 f rom 1:00-4:00 p.m.

A list of attendees for this meeting is attached (Enclosure 1).

The following items were addressed at the meeting:

1 1.

The Committee continued at this meeting their review (begun at Meeting No.167) of the proposed (comoined) resolution for Generic Issue 70 (Power-0 pet < W Relief Valve and Block Valve Reliability) and Generic Issue 94 (In tlonal Low-Temperature Overpressure Protection for Light Water Reactt,rs).

The Committee recommended in favor of issuing the proposed resolution and implementing generic letter, subject to a number of minor modifications (to be coordinated with the'CRGR staff).

This l

matter is discussed in Enclosure 2.

2.

The Committee was briefed on issues to be considered in the development of guidance for the treatment of external events in the.IPE, and milestones of the projected schedule for the development of that-guidance.

The Committee made no recommendations with regard to IPEEE guidance at this meeting.

These matters are discussed in Enclosure 3.

In accordance with the ED0's July 18, 1983 directive concerning " Feedback and Closure of CRGR Reviews," a written response is required from the cognizant office to report agreement or disagreement with the CRGR recommendations in these minutes.

Tha response, which is required within five working davs after receipt of these minutes, is to be forwarded to the CRGR Chairman and if there is disagreement with CRGR recommendations, to the EDO for decisionmaking.

Questions concerning these meeting minutes should be referred to Jim Conran (492-9855).

Original S!gr,so Sy' O. J. Heltemas, Jr.

v Edward L. Jordan, Chairman Committee to Review Generic Requirements

Enclosures:

Distribution: (w/o enc.)

As stated Central File CRGR SF (w/ enc.)

PDR (NRC/CRGR)

M. Taylor (w/ enc.)

cc:

See next page S. Treby.

L. Shao (w/ enc.)

W. Little R. W. Houston (w/ enc.)

J. Conran (w/ene.

M. Lesar R. Baer (w/ enc.)

D. Allison (w/ enc.)

P. Kadambi (w/ enc.)

E. Jordan (w/ enc.)

CRGR CF (w/ enc.)

J. Heltemes (w/ enc.)

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Material Related to CRGR Meeting No.168 Io Be Made PublicIv Available 1.

Memo dated September 6,1989 for J. Taylor from E. Jordan,

Subject:

Minutes of CRGR Meeting Number 168, including 2 enclosures which were not previously released:

a., a summary of discussions of a proposed resolution for GI-70 on PORY and block valve reliability and GI-94 on additional i

LTOP for LWR's, including I attachment.

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b., a summary of a briefing on development of guidance for IPEEE, including 2 attachments.

SENT TO PDR ON 7/30/90 l

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o ATTENDANCE LIST FOR CRGR MEETING NO. 168 August 23, 1989 CRGR MEMBERS E. Jordan D. Ross J. Goldberg G. Ar-lotto L. Reyes J. Sniezek l

CRGR STAFF J. Heltemes C. Sakenas J. Conran D. Allison NRC STAFF E. Throm G. Mazetis F. Cherny R. Kirkwood R. Baer K. Kniel W. Minners C. Liang J. Lyash L. Plisco J.-Chen l

D. Jeng W. Beckner N. Chokshi l

L. Shao

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G. Kelly A. Murphy L. Reiter C. McCracken G. Bagchi J. Mitchell G. Cummings

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. to the Minutes of CRGR Meeting No. 168 Proposed (Combined) Resolution for GI-70 and GI-94 August 23, 1989 TOPIC The Committee continued at this meeting their review (begun at Meeting No.

167) of the combined resolution for GI-70, " Power-0perated Relief Valve and Block Valve Reliability" and GI-94, " Additional-Low-Temperature Overpressure Protect on for Light Water Reactors." The presentation for the GI-70 portion of the combined resolution package was completed at Meeting No. 167; E. Throm (RES) presented the GI-94 portion of the combined resolution at this meeting.

Copies of the briefing slides used for both GI-70 and GI-94 are attached to the minutes of CRGR Meeting No. 167.

BACKGROUND 1.

The documents submitted initially for CRGR-review in this matter are identified in the the minutes for Meeting No. 167.

2.

The staff provided directly to CRGR members at this meeting the following additional / revised documents:

Proposed Generic letter (undated) regarding GI-70 and GI-94 combined a.

resolution, revised and updated to reflect CRGR comments in Meeting No. 167.

(Attachment 1 to this Enclosure)

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b.

Memorandum dated August 22, 1989, R. Baer/W. Houston to E. L. Jordan, transmitting analysis (requested by CRGR at Meeting No. 167) of the risks and benefits of keeping AC power available to PORV block valve.

(Attachment 2 to this Enclosure)-

CONCLUSIONS / RECOMMENDATIONS As a result of their. review of this matter, including the discussions with the staff at this meeting and Meeting No. 167, the Committee recommended in favor of issuance of the combined resolution for GI-70 and GI-94 proposed by the staff, subject to the following modifications that are to be coordinated with the CRGR staff:

NOTE:

ALL THE RECOMMENDATIONS BELOW ARE KEYED TO THE REVISED VERSION O THE GENERIC LETTER DISTRIBUTED BY RES AT MEETING NO. 168

.(I.E., BACKGROUND ITEM 2.a ABOVE)

1.

Recommended Modifications to Proposed Generic Letter a

a.

Page 1 i.

Change addressee line to read as follows:

"To All PWR Licensees and CP Holders" Make conforming changes throughout the package as required.

ii.

End the first sentence of the first paragraph after the words

"... Enclosures A and B of this letter."

iii. Delete all but the last sentence of the second paragraph.

Move the last sentence to the end of.the first paragraph, b.

Page 2 i.

End the second to last sentence of the first paragraph after the words "... justification for.any delay."

ii.

Delete the last sentence of the first paragraph.

iii. Change the wording immediately following the first paragraph to read as follows:

"The response to this letter shall include the following specific items."

iv.

Delete response items 1 and 5.

c.

Psge 3 i.

Delete the first sentence of the second paragraph.

ii.

In accordance with guidance provided by the CRGR staff, include a brief section in the generic letter that (a) clearly states that the actions requested by this' generic letter are cost-beneficial backfits that enhance significantly, and (b) briefly summarizes the justification for the backfits involved.

2.

Recommended Modifications to Appendix A of the Generic Letter a.

Page A-2 i.

In the last lir.e of the second paragraph, change the word

" safety" to " safety-related." Check throughout the package for proper use of the term safety-related (as the term is defined in 10 CFR 50.49 and Appendix A to 10 CFR Part 100).

ii.

Delete the next to last sentence in the third paragraph.

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-3 b.

Page A-3 i.

Reword paragraph 3.1.(c) to indicate more clearly.that for nonsafety grade replacement parts, spares, etc., the staff's intent is to neither upgrade nor downgrade the current quality classification..

c.

Page A-5 1.

Under Action a, in'the second line, insert the following phrase af ter the words "... restore the PORV(s)...":

"...with power maintained to the block valve (s)..."

3.

Recommended Modifications to Appendix B of the Generic Letter a.

- Page B-1 1.

Change the first sentence of the second paragraph to read as follows:

"... designated as Generic Safety Issue A-26...and partially resolved in 1979 by..."

ii.

Change the second sentence of the second paragraph to read as follows:-

"PWR licensees implemented procedures in response to a generic letter to reduce the potential..."

b.

Page B-2 i.

For clarity and consistency in the package, in the third line of the fourth paragraph, insert the term "PWR" between the words all and " plants."

i 11.

In the fifth and sixth paragraphs, the word " requirements" should be changed to " guidance" or " positions" or

" recommendations." The staff should re review the wording throughout the rest of the package for proper usa terms, and make conforming changes as necessary, ge of these' c.

Page B-3 i.

In the first full paragraph, delete the next to last sentence (and associated handwritten revisions).

ii.

Delete the next to last paragraph.

d.

Page B-5 i.

Conform the wording of the second full paragraph to the wording of Action Statements b and c on page 8-6, after making the changes to those Action Statements recommended in 3.e below.

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i 11.

The last paragraph is out of place and in the wrong context where it is currently located.

It really refers to staff I

assumptions on which T/S Action Statements are based.

The staf f should consider locating this information either in.T/S Bases or in the Generic Letter itself.

e.

Page B-6

'i.

Change Action Statement b-to read as follows:

"With one PORV inoperable in Modes 5 or 6, (1) restore the inoperable PORV to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or (2) depressurize-and vent the RCS through at least a-square inch vent within 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />."

11.

Change Action Statement c to read as follows:

"With both PORVs inoperable, complete depressuri-zation of the RCS through at least a square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />."

111. The Committee-recommended that the changes in 3.e.i and 3.e.ii above be reflected.also in the Standard Technical Specifications maintained by NRR.

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I To All Pressurized Water Reactor Licensees and Applicants

SUBJECT:

NRC Position on the Resolution of Generic issue 70,

" Power-0perated Relief Valve and Block Valve Reliability" and Generic Issue 94, " Additional Low-Temperature Overpressure Protection for Light-Water Reactors" (Generic Letter 89-xx)

The purpose of this Generic-Letter is to advise pressurized water reactor (PWR) licensees and applicants of the staff positions delineated in Enclosures A and

.B to this letter and to determine whether licensees and applicants will voluntarily agree to adopt these positions.

Enclosure A presents the staff position resulting I

from the resolution of GI-70 and is applicable to all Westinghouse and Babcock and Wilcox (B&W) designed plants, and Combustion Engineering (CE) designed plants with PORVs. Enclosure B presents the staff position resulting from the resolution of GI-94 and is applicable to all Westinghouse and CE designed plants whether or not they have FORVs and block valves.

This Generic Letter applies to all PORVs that are used or could be used'to perform one or more of the safety-related functions identified in Section 2 of Enclosure A whether or not credit for use of PORVs is specifically included in a plant's licensing bases documents, such as a Final Safety-Analysis Report.

i Enclosure A does not apply to-those CE designed plants that do not have PORVs and block valves.

Enclosure B does not apply to B&W designed plants.

On the basis of technical studies for GI-70, the' staff requests that to enhance safety, actions identified-in Section 3 of Enclosure A be taken by all l

PWRlicenseesandappJl pat use or could use PORVs to perform any of the safety-related function 3

. Tion 2 of Enclosure A.

These actions result from the steff interpretation of C=:=1 Od:i;n Crit:ri; M ad 0: in '0CiZ C, nppx;i

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So AUD 10 6fC PAKT (00, k RNOlx A )f2TY~RfkATED ArQVIfMBM (EZE 10 ff On the basis of technical studies for GI-94, the staff also requests that to enhance safety, actions identified in Section 3 of Enclosure B be taken by all Combustion Engineering ar.d Westinghouse PWR licensees and applicants. These actions result from the staff interpretation of General Design Criteria 15.and 31,inIQC,F 0, Appendix A.

The information requested by this letter is directed a ddressing these concerns.

Pse.r Note that the staff's requests are based on the performance of PORV and:

PORV block valve designs used to date on:U.S. power reactors.

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The use of more reliable valves should result in less f_regp_ent corrective traintenance and can result in longer inservice testing intervals as delineated in Section XI'of.the ASME Boiler and Pressure Vessel Code.

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Accordingly, pursuant to'Section 182 of the Atomic Energy Act and'10 CFR 50.54 (f), you, as a PWR licensee or applicant, are required to advise the NRC staff under oath or af fimation, within 180 days of the date of this letter, of your f

current plans relating to PORVs and block valves and to low-temperature over-l pressure protection, in particular whether you intend to follow the staff positions ir.cluded in Enclosures A and B as applicable, attached to this letter, I

orproposealternatemeasures,andyourproposedscheduleforimplementation.u.d staf f positiensgshould be implemented by.the end of the first refueling outage that,t;.. _ ::- the date of this letter.

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roposed revis8' A any delay, ano any planned compensating safety actions to be taken during the interim.

The staff will evaluate the licensee's or applicant's response and

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An acceptable res'ponse to this letter would include the following specific items.

1.

A list identifying those safety-related functions for which the PORVs are used. (Section 2 of Enclosure A that my be performed by PORVs.){dentifies those safety-related functions STAYg b uD 3 A40 APPueMTs A ___ggMy, improvements 1[t'm;h ' in Section 3.1 of Enclosure A._ _ by lic 2.

With respect to improvement (3,1n Section 3.1 of Enclosure A, licensees should state whether they will comrait to use those modified limiting conditions of operation of PORVs and block valves in the technical specifications 4%) for Modes 1,.

2, and 3 in Attachment A-1 of Enclosure A for Westinghouse and CE designed plants with two PORVs, or in Attachment A-2 of Enclosure A for Westinghouse designed plants with three PORVs, or in Attachment A-4 of Enclosure A for B & W d Q1np lants #

In addition to this 10 CFR 50.54(f) request, if the literrfeel

..dN8implementthesereg requested that they submit modifications to dg r g;___.. echnical-" r: -Q tggjyf o T

1. PK4//rfc4fied specifications in a license amendment r:NCHROUkk N0YKO A60VEo.;c m unt, n r: Of th:

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MD APPL 64 Nit,W SPIECim C ATIONS A s ta tement by licensees aslC A,4TSto whether they wi submit a liceMe ambn~dment S #.

s request to modify the Technical Specification for the low-temperature overpressure protection system concerning the imiting conditions of operation in Modes 5 and 6 as identified in Attachment B-1 of Enclosure B to this Generic Letter for Westinghouse or CE designed plants, as appropriatej AWD APPL lCANTS 4 F.

A sta te ent by licer.seesg s to wnether they will revise plant cooldown and a

hea tup p-ocedures to incorporate the revised limiting conditions for operation.

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INSERT 1 The Technical Specification modifications in staff position-3 in Section 3.1 of Enclosure A and in Section 3 of Enclosure B should be submitted by the end of the first refueling outage that starts six months or later from the date of this letter.

For PWR Construction Permit Holders, staff positions 1 and 2 in Section 3.1 of Enclosure A should be implemented prior to initial criticality or within six months of the date of the letter, whichever is later.

The Technical Specification modifications in staff position 3 in Section 3.1 of Enclosure A and in Section 3 of Enclosure D should be submitted by the end of the first refueling outage that starts six months or-later from the date of this letter.

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-The actions to incorporate TS requirements for the resolution of GI-70 and GI-94 are considered to be consistent with the Comission's Policy Statement on Technical Specification Improvements.

That policy captures existing requirements under Criterion 3 (Mitigation of Design Basis ~ Accidents or Transients) or under the provisions to-retain requirements which operating experience and probabilistic risk assessment show to be important to public health and safety. While it is recognized that PORVs for older plants may not have been classified as safety-related components that are used to mitigate a design basis accident and, thereby, they may not have been included in TS,as part of the plant's licensing basis, this is not an acceptable basis for not implementing the proposed actions to incorporate TS requirements for PORVs consistent with the guidance provided. Likewise, such requirements would oe retained in TS when implementing improvements in TS consistent with_the Commission Policy Statement on the basis of Criterion 3.or risk considerations noted above.

-The staff will review responses to this Generic Letter to assure that appropriate-actions have been or will be taken, on an appropriate schedule, and that TS are in place for operating plants or will be included in TS for operating license applications that provide an acceptable resolution for the concerns identified by GI-70 and GI-94.

Alternatives to schedules and the guidance provided herein will be evaluated'on their merits on an individual case basis. Based on its review and the acceptability of these responses', the staff will close out GI-70 and GI-94 for each plant.

This request is covered by Office of Management and Budget Clearance Number 3150-001 which expires December 31, 1989.

The estimated average burden hours is 320 man-hours per licensee response, including assessment of the new recomendations, searching. data sources, gathering and analyzing the data, and preparing the required reports.

Coments on the accuracy of this estimate and suggestions to reduce the burden may be directed to the Office of Management and Budget, Room 3208, New Executive Office Building, Washington, D.C.

20503, and the U.S. Nuclear Regulatory Comission, Records and Reports Management Branch,' Office of Administration-and Resources Management, Washington, D.C.

20555.

Sincerely, James G. Partlow Associate Director for Projects Office of Nuclear Reactor Regulation h)

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Enclosures:

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Enclosure A-Staff Positions Resulting from Resolution of Generic Issue 70.

Enclosure B-Staff Positions Resulting from Resolution of Generic Issue,94.

Enclosure C-NUREG-1316, " Evaluation of Power-Operated Relief Valve and Block Valve Reliability in PWR Nuclear-Power Plants."

Enclosure D-NUREG-1326, " Regulatory Analysis for the Resolution of Generic Issue 94, Additional Low-Temperature Overpressure Protection for Light-Water Reactors."

1 Plants that already have identified safety-related functions during the licensing process, and have a staff' safety evaluation report that appruves the 5

identified functions need merely state this.in their response.

No further action will be required for this aspect of the Commission's position.

2 Plants that already have staff _ issued technical specifications consistent with these requirements need merely state this in their response.

No further action will be required for this aspect of the Commission's position.

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Enclosure A to Generic Letter 89-XX Staff Positions Resulting from Resolution of Generic Issue 70 -

PORV and Block Valve Reliability 1.

BACKGROUND 1

Generic Issue 70 (GI-70), " Power-0perated Relief Valve and Block Valve Reliability,"-involves the evaluation _ of the reliability of' power-operated relief valves (PORVs) and block valves and their safety significance in PWR plants.

The technical findings and regulatory analysis related to GI-70 are discussed in NUREG-1316. " Evaluation of Power-Operated Relief Velve and Block Valve Reliability'in PWR Nuclear Power Plantt," enclosed.

This report identifies those safety related functions that may be performed by PORVs and also identifies potential improvements to PORVs and block valves.

In support of the resolution of GI-70, the Oak Ridge National Laboratory-(ORNL) performed a study of PORV and block valve operating experience.

A report, prepared by ORNL, was issued,as'NUREG/CR-4692, " Operating Experience Review of Failure of Power Operated Relief Valves and Block Valves in Nuclear Power Plants," dated October 1987.

Traditionally, the PORV and its block valve are provided for plant operational flexibility and for limiting the number of challenges to the safety-related pressurizer safety valves.

The operation of the PORVs has not been classified as a safety-related function,' i.e., one on which the results and conclusions of the safety analysis is based and that invokes the highest level of quality

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and construction.

For overpressure protection of the reactor coolant pressure boundary (RCPB) at normal operating temperature and pressure, the operation-of PORVs has not been explicitly considered as a safety-related' function.

Also, an inadvertent opening of a-PORV or safety valva has been analyzed in the Final

' Safety Analysis Reports as an anticipated operational occurrence with acceptable l

consequences.

For these reasons, most PWRs, particularly those licensed prior to 1979, do not classify PORVs as safety-related components.

l The Three Mile Island Unit 2 (THI-2) accident focused attention on the reliability of PORVs and block valves since the malfunction of the PORV at TMI-2 contributed to the severity of the accident.

On other occasions PORVs have stuck open when called upon to function.

Also, there are PORVs in many operating plants that have leakage problems, so that the plants must be operated with the upstream block valves in the closed position.

The technical specifications governing L

PORVs on most operating PWRs, that deal with closing the block valve and removing i

power, were developed to allow continued plant operation with degraded PORVs, but did not consider.the need for the PORVs to perform the safety functions discussed below.

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l Following the TMI-2 accident, the staff began to examine transient and accident events in more detail, particularly with respect to required operator actions and equipment availability and performance.

As a result, the staff initiated l

an evaluation of the role of PORVs to perform certain safety related functions, j

I A-2 2.

SAFETY FUNCTIONS OF PORVs AND BLOCK VALVES i

l The staf f, in its evaluation, ' determined that over a period of time the role of PORVs has changed such that PORVs are now relied upon by man B&W, and CE designed plants with PORVs to perform one, or more,y Westinghouse, safety related functions:

of the following 1.

Mitigating a design-basis steam generator tube rupture accident, i

2.

Low temperature overpressure protection of the reactor vessel during-startup and shutdown,et I

3.

' Plant cooldown in compliance with Branch Technical Position RSB 5 to SRP 5.4.7, " Residual Heat Removal (RHR) System", won 4

c:nting.

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i Where PORVs are used or could be used to perform one, or more, of the 1

safety related functions identified above, or to perform any other safety related function that may be identified in the future, it'is appropriate to reconsider 1

the' safety classification of PORVs and the associated block valves.

i plants receiving an operating license in recent years, the ~ staff has required For certain these valves to be classified as safety related components if they perform one or more, safety functions.

PWR For operatingAplants the staff has' concluded that it is not cost effective to replace (backfit) existing nonsafety grade PORVs.and block valves (and associated control systems) with PORVs.and block valves that are safety grade !

even when they have been determined to perform any of the safety-related functions discussed above.

Subsequent to the TM1-2 accident, a number of improvements were required of PORVs and block valves,'such'as requirements to be powered from Class IE buses and to have valve position indication in the control room.

Therefore, the additional benefits that would result from-upgrading PORVs to full safety grade status are. considered to be of marginal benefit.

For operating plants, the greatest immediate benefits can be derived i

from implementing items 1 through gnidentified below which can increase the reliability of these components and provide assurance they will function as required.

1 1

3.

lHPROVEMENTS TO ALL PORVs AND BLOCK VALVES 3.1 Operatino PWR Plants we OcuSTE.UC#tou Peas Hos.ozes Based on the analysis and findings for GI-70, the staff concludes that the following actions should be taken to improve the reliability of PORVs and block valves:

1.

Include PORVs and block valves within the scope of 44wk operational AW quality assurance program.that is in compliance with 10 CFRg50, Appendix B.

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Include PORVs, valves in PORY control systems, and block valves within the scope of a program covered by Subsection IWV, " Inservice Testing of Valves in Nuclear Power Plants," of Section XI of ASME Boiler and Pressure Vessel Code.

As pennitted by the Code. Stroke testing of PORVs should only be performed during[ mode 4 (hot shutdown) and in all cases prior to establishing conditions where the PORVs are used for low-temperature overpressure protection. Stroke testing of the PORVs should not be performed during power operation. Additionally, the PORV block valves should be included in the-licensees' response to the expanded MOV test program discussed in NRC Generic Letter 89-M/o "Safet{-RelatedMotorOperatedValveTestingandSurveillancedated (1:t=,34K St,lf tf.

ants modify the limiting conditions of operation of 3 #,

For operating $k valves in the technical specifications for Modes 1, PORVs and blo l

2, and 3 to incorporate the position adopted by the staff in recent i

t licensing actions. Attachments A-1 through A-3 are provided fo-pwg guidance.

The staff recognizes that some recently licensed ants already have technical' specifications in accordance with the staff position. Such plants are already in compliance with'this position and need merely to state that in their response. These recent technical specifications require that plants that 'run with the block valves closed, (e.g., due to leaking PORVs) maintain electrical power to the block valves so they can be readily opened from the control room upon demand.

Additionally, plant operation in Modes 1, 2, and 3 with PORVs and block valves inoperable for reasons other than seat leakage is not permitted for periods of more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

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A-5 Generic Issue 70 Enclosure A to Generic Letter 89-XX Attachment A-1 Modified Standard Technical Specifications for Combustion Engineering and Westinghouse Plants REACTOR COOLANT SYSTEM 3/4.4.4 PELIEF VALVES LIMITING CONDITION FOR OPERATION t

The following is to be used when.two PORVs are provided:

3.4.4 Both power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With one or both PORVs inoperable, because of excessive seat a.-

leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV s) to OPERABLE status or close the associated block valve (s);(otherwise, be in~ at least HOT STANDBY within the next 6' hours and in HOT.SHUTr)WN the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, u

b.

With one' PORY inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to 0PERABLE status or close its associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following l

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT.

SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With both PORVs-inoperable due to causes other than excessive seat c.

leakage, within I hour either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, d.

With one or both block valves inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the.

block valve (s) to OPERABLE status or place its associated PORV(s) in manual control.

Restore at least one block valve to OPERABLE status within the next hour if both block valves are inoperable, restore any remaining inoperable block valve to operable status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise be in-at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The provisions of Specification 3.0.4 are not applicable..

e.

i

,=

A-6 SURVEILLANCE REQUIREMENTS 4.4.4.1 In. addition to the requirements of Specification 4.0.5, each PORY shall be demonstrated OPERABLE at least once per 18 months'by:

Operating the PORV through.one complete cycle of full travel'during a.

"O b'

,1n, ope b.

Where applicable, operating solenoid air control valves and check valves on associated air accumulators in PORY control systems through one complete cycle of full travel for plants with air-operated PORVs, and Performing a CHANNEL CALIBRATION of the actuation instrumentation.

c.

4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days.by operating the valve through one complete cycle of full travel, unless c in Specification 3.4.4the block valve is closed in order to meet the requirements of A 4.4.4.3 The emergency power supply for the PORVs and block valves shall.be demonstrated OPERABLE at least once per 18 months by:

Manually transferring mot'.ve and control power from the normal to a.

the emergency power bus, and b.

Operating the valve', through a complete cycle of full travel.

COMBUSTION ENGINEERING AND WESTINGHOUSE PLANTS I

I

___-_a

.1 A-7 Generic Issue 70

- Enclosure A To Generic Letter 89-XX

-Attachment A-2' Modified Standard Technical Specifications for Westinghouse Plants with Three PORVs REACTOR COOLANT SYSTEM' 3/4.4.4 REllEF VALVES LIMITING CONDITION FOR OPERATION The.following is to be used when three PORVs are provided:

3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

With one or more PORVs inoperable because'of excessive seat leakage,-

a.

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s): to OPERABLE status or close the~ associated block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, b.

With one or two PORVs inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or close the associated block-valve (s) and remove power from the block valve (s); restore the PORV(s) to OPERABLE status within the following 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />slor be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, With three PORVs inoperable due to causes other than excessive seat c.

leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore at least one PORV to OPERABLE status or close the block valves and remove power from the block valve (s) and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d.

With one or more block valves inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the block valve (s) to OPERABLE status or place its associated PORV in manual control.

Restore at least one block valve to OPERABLE status witn..i the next hour if three block valves are inoperable, restore any remaining inoperable block valve (s) to operable status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The provisions of Specification 3.0.4 are not applicable.

e.

A-8 SURVEILLANCE REQUIREMENTS i

4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORY shall be demonstrated OPERABLE at least once per 18 months by:

a.

Operating the PORV through one complete cycle of full travel during Modes Q nd

'b, 0 9R.

b.

Where applicable, operating solenoid air control valves and check valves on associated air accumulators in PORV control systems through one complete cycle of full travel for plants with air-operated PORVs, and c.

Perfonning a CHANNEL CALIBRATION of the actuation instrumentation.

4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION a, b, or c in Specification 3.4.4.

4.4.4.3 The emergency power supply for the PORVs and block valves shall be demonstrated OPERABLE.at least once per 18 months by:

a.

Manually transferring motive and control power from the normal to the emergency power bus, and b.

Operi ting the valves through _a complete cycle of full travel.

WESTINGHOUSE PLANTS

. - - = _ _.

i; A-9 Generic Issue 70 Enclosure A to Generic Letter 89-XX Attachment A-3 Applicable To Combustion Engineering and Westinghouse Plants 3/4.4.4 RELIEF VALVES i

Bases of the Limiting Condition for Operation (LCO) and Surveillance Reauirements:

The OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following functions:

A.

Manual c'ontrol of'PORVs to control reactor coolant system pressure.

This is a function'that is used for the steam generator tube rupture accident and for plant shutdown.

This function has been classified as safety related for more recent plant designs.

B.

Maintaining the integrity of the reactor coolant pressure boundary.

This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.

C.

Manual control of_ the block valve to:

(1) unblock an isolated PORY te allow it to be used for manual control of reactor coolant system pressure (Item A), and (2) isolate a PORV with excessive seat leakage (Item B).

D.

Automatic control of PORVs to control reactor coolant system pressure..

This is a. function that reduces challenges to the code safety valves for overpressurization events.

l E.

Manual control of a block valve to isolate a stuck open PORV.

L Surveillance Requirements provide the assurance that the PORVs and block valves can perform their functions.

Specification 4.4.4.1 addrerses PORVs, l.

'4.4.4.2 the block' valves, and 4.4.4.3 the emergency _(backup) power sources.

The latter are provided for either PORVs or block valves, generally as a consequence of the TMI action requirements to upgrade the operability of-PORVs-and block valves, where they are installed with non-safety grade power sources, including instrument air, and are provided with a backup-(emergency) power source. The block valves are exempt from the surveiliance requirements to cycle the valves when they have been closed to comply with the ACTION l-requirements. This precludes the need to cycle the valves with full system l

differential pressure or when maintenance is being performed to restore an inoperable PORY to operable status.

1 Surveillance requirement 4.4.4.1.b has been added to include testing of the mechanical and electrical aspects of control systets for air-operated PORVs.

l

A-10 rot STAOD5Y ore.

3 Testing of PORVs in HOT SHUTDOWN is required in order to simulate ~ the temperature and pressure environmental effects on PORVs.

In many PORY designs, testing at COLD SHUTDOWN is not considered to be a representative test for assessing PORV performance under normal plant operating conditions.

The Modified Standard Specification (STS) requirements include the following changes from prior STS guidance:

1.

Clarify the statement of LCO by replacing "All" with "Both" where the design includes two PORVs.

2.

ACTION statement a. does not include the requirement to remove power from closed block valve (s), because removal of power would render the block valve (s) inoperable and the requirements of ACTION c, would apply.

Power would be maintained to the block valve (s) so that it is operable and may be subsequently opened to allow the PORY to be used to control reactor pressure.

Closure of the block valve (s) establishes reactor ccolant pressure boundary integrity for a PORV that has excessive seat leakage.

(Reactor coolant pressure boundary integrity takes priority over the capability of the PORY to mitigate an overpressure event.)

3.

ACTION statements b. and c. includes the removal of power from a closed block valve as additional assurance to preclude any inadvertent opening of the block valve at a time in which the PORV may not be closed due to maintenance to restore it to OPERABLE status.

(In contrast, ACTION statement a permits continued plant operation with the block valves closed' i.e.

continued operation is not dependent on maintenance to eliminate excessive PORV leakage, and, therefore, ACTION statement a. does not require removal of power from the blockvalve.)

i 4

ACTION statements a.,

b., and c. have been changed to tenninate the forced shutdown requirements with the plant being in HOT SHUTDOWN rather than COLD SHUTDOWN because the APPLICABILITY requirements of the LCO do not extend past the HOT STANDBY Mode.

5.

ACTION statement d. has been modified to establish remedial measures that l

are consistent with the function of the block valves.

The prime importance for the capability to close the block valve is to isolate a stuck open PORV. Therefore, if the block valve (s) cannot be restored to operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the remedial action is to place the PORV in manual control to preclude its automatic opening for an overpressure event and to avoid the potential for a stuck open PORY at a time that the block valve is inoperable. The time allowed to restore the block valve (s) to operable status is based upon the remedial action time limits for inoperable PORVs per ACTIONS statements b. and c. since the PORVs are not capable of mitigating an overpressure event when placed in manual control. These actions are also consistent with the use of the PORVs to control reactor coolant system pressure if the block valves are inoperable at a time that they have been closed to isolate PORVs that have excessive seat leakage.

The modified ACTION statement does not specify closure of the block valves because such action would not likely be possible when the block valve is inoperable. Likewise, it does not specify either the closure of the PORV,

A-11 because it would not likely be open, or the removal of power from the PORV.

When the block valve in inoperable, placing the PORY in manual control is sufficient to preclude the potential for having a stuck open PORV that could not be isolated due to an inoperable block valve.

For the same reasons, reference is not made to ACTION statements b. and c. for the required remedial actions.

Surveillance requirement 4.4.4.2 has been modified to remove the exception-6.

for testing the block valves when they are closed to isolate' an inoperable PORV.

If the block valve is closed to isolate a PORY with excessive seat leakage, the operability of the block valve is of importance, because opening the block valve is necessary to permit the PORV to be used for manual control of reactor pressure.

If the block valve is closed to isolate an otherwise inoperable PORV, the maximum allowable outage time is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> which is well within the allowable limits (25 percent) to extend the block valve surveillance interval (92 days).

Furthermore, these test requirements would be completed by the reopening of a recently closed block valve upon restoration of the PORY to operable status, i.e., completion of the ACTION statement fulfills the required surveillance requirement.

1 I

l i

4 e

w

i A-12 l

Gener'ic Issue 70 Enclosure A to Generic Letter _89-XX i

Attachment A-4 Modified Technical Specifications for Babcock and Wilcox Plant REACTOR COOLANT SYSTEM-3/4.4.4 REllEF VALVE

. LIMITING CONDITION FOR OPERATION 3.4.4 The power-operated relief valve (PORV) and its-associated block valve shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION.

With' the PORY inoperable - because of excessive seat leakage, within I a.

hour either restore-the PORV to OPERABLE status or close the associated block valve; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With the PORY inoperable due to causes other than excessive seat i

i leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve, i

and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With the block valve inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore.the block c.

valves to OPERABLE status or place the associated PORY-in manual control and restore the block valve to operable status within the next hour; otherwise be in HOT STANDBY within'the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN in the following.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d.

The provisions of Specification 3.0.4 are not applicable.

_ SURVEILLANCE REQUIREMENTS 19 Apolviou ToTHE ReaviteHEMTS OF SpactrickTion 4.o.5 4.4.4.lgthePORVshallbedemonstratedOPERABLEatleastonceper18 months 3

by:

Operating the PORY through one complete cycle of full travel during a.

Modes, and

3) o R:,

A-13 b.

Performing a CHANNEL CAllBRATION of the actuation instrumentation.

4.4.4.2 The block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION a, or b in Specification 3.4.4.

4.4.4.3 The emergency power supply for the PORV and block valve shall be demonstrated OPERABLE at least once per 18 months by:

a.

Manually transferring motive and control power from the normal to l

the emergency power supply, and b.

Operating the valvef through a complete cycle of full travel.

BABCOCK & WILCOX-PLANTS l

1 l

'l A-14 Generic Issue 70

, Enclosure A to Generic Letter 89-XX Attachment A-5 Applicable To Babcock'and Wilcox. Plants 3/4.4.4 RELIEF VALVE Bases of the Limiting Condition for Operation (LCO) and Surveillance Requirements:

The OPERABILITY of the PORY and block valve is determined on the basis of their being capable of performing the following functions:

A.

Manual control of the PORV to control reactor coolant system pressure.

This is a function that is used for the steam generator tube rupture accident and for plant shutdown.. This function has been classified as~

safety-related.for more recent plant designs.

B.

Maintaining the integrity of the reat;or coolant pressure boundary. This is a function that is related to cont olling. identified leakage and 4

ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.

i C.

Manual control of the block valve to:

(1)unblockanisolatedPORYto allow it to be used for manual control of reactor coolant system pressure (Item A), and (2) isolate the PORY with excessive seat leakage-(Item B).

D.

Automatic control of the PORV to control reactor coolant system pressure.

This is a function that reduces challenges to the ; ode safety valves for l.

overpressurization events.

i E.

Manual control of a block valve to isolate a stuck open PORV.

i Surveillance Requirements provide the assurance that the PORV and block valve can perform their functions.

Specification 4.4.4.1' addresses the PORV, 4.4.4.2 the block valve, and 4.4.4.3 the emergency (backup) power source.

The latter is provided for either the PORV or block valve, generally as a consequence of the TMI action requirements to upgrade the operability of PORVs and block valves, where they are installed with non-safety grade power sources, including instrument air, and are prov.ded with backup (emergency-power sources.

The block valve is exempt from the surveillance requirements to cycle the_ valve when it has been closed to comply with the ACTION requirements.

This precludes the need to cycle the valve with full system differential pressure or when maintenance is being performed to restore an l

inoperable PORV to operable status.

l l

!~

A-15 W OT STA.0 DRY opll:

Testing the PORV inhT SHUTDOWN is required in order to simulate the te 3

and pressure environmental effects on the PORY.

at COLD SHUT 00WN is not considered to be a representative test for asse PORY performance under normal plant operating conditions.

The Modified Standard Specification (STS) requirements include the following changes from prior STS guidance:

1.

ACTION statement a. does not include the requirement to remove power from the closed block valve, because removal of power would render the block valve inoperable and the requirements of ACTION c. would apply.

Power would be maintained to the block valve so that it is operable and may be subsequer.tly opened to allow the PORY to be used to control reactor pressure.

Closure of the block valve establishes reactor coolant pressure boundary integrity for a PORV that has excessive seat leakage.

integrity takes priority over the capabilif(Reactor coolant pressure boundary

/ of the PORY to mitigate an overpressureevent.)

2.

ACTION statements b. includes the removal of power from the closed block i

valve as additional assurance to preclude any inadvertent opening of the block va?ve at c time in which the PORV may not be closed due to maintenance to rci we it to OPERABLE status.

(In contrast, ACTION statement a. pennits cononued plant operation with the block valve closed, i.e., continued operation is not dependent on maintenance to eliminate excessive PORY leakage, and, theretore. AC"10N statement a. does not require removal-of power from the block valve.)

ACTION statements 'a and b. have been changed to terminate the forced shutdown 3.

requirements with the plant being in HOT SHUTDOWN rather than COLO SHUTDOWN because the APPLICABILITY requirements of the 1.00 do not extend past the HOT STANDBY Mode.

ACTION statement c. has been modified to establish remedial measures th are consistent with the function of the block valves.

The prime importance for the capability to close the block valve is to isolate a stuck open PORV.

Therefore, if the block valve cannot be restored to operabia status within I hour, the remedial action is to place the PORY in manual control to preclude its opening for an overpressure event and to avoid the open PORY at a time that the block valve is inoperable. potential for a stuck The time allowed to restore the block valve to operable status is based upon the remedial action time limits for inoperable PORV per ACTION staternts b. Since the PORY is not capable of mitigating an overpressure event when placed in manual control.

This action is also consistent with the use of the PORY to control reactor coolant system pressure if the block valve is inoperable at 3 time that it was closed tc isolate a PORY that has excessive seat leakage.

The modified ACTION statement does not specify closure of the block valve because such action would not likely be possible when the block valve is inoperable.

Likewise, it does not specify either the closure of the PORV, because it would not likely be open, or the removal of power from the PORV.

When the block valve is inoperable, placing the PORY in manual control is sufficient to preclude the potential for having a stuck open PORV that could not be isolated due to an inoperable block valve.

For the same reasons, reference is not made to ACTION statement b. for the required remedial action.

A-16 5.

Surveillance requirement 4.4.4.2 has been modified to remove the exception for testing the block valve when it is closed to isolate an inoperable PORY.

If the block valve is closed to isolate a PORV with excessive seat leakage, the operability of the block valve is of importance, because opening the block valve is necessary to permit the PORY to be used for manual control of reactor pressure.

If the block valve is closed to isolate an otherwise inoperable PORY the maximum allowable outage time is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> which is well within the allowable limits (25 percent) to extend the block valve surveillance interval (92 days).

Furthermore, these test requirements would be completed by the reopening of a recently closed block valve upon restoration of the PORY to operable status, i.e., completion of the ACTION statement fulfills the required surveillance requirement.

i l

Enclosure B to Generic Letter 89-XX Staff Positions Resulting from Resolution of Generic Issue 94 -

Additional Low-Temperature Overpressure Protection For Light-Water Reactors

1. BACKGRbUND Generic Issue 94 (GI-94), " Additional Low-Temperature Overpressure Protection for Light-Water Reactors," addresses concerns with the implementation of the requirements set forth in the resolution of Generic Safety Issue A-26. " Reactor Vessel Pressure Transient Protection (Overpressure Protection)." In support of GI-94, the Battelle Pacific Northwest Laboratory (PNL) performed a study based on actual operating reactors experiences to determine the risks associated with current low-temperature overpressure protection (LTOP) systems.

A report, prepared by PNL, has been issued as NUREG/CR-5186, " Low Temperature Over-pressure Systems Evaluation for Pressurized Water Reactors," dated November 1988.

The staf f has prepared a Regulatory Analysis for GI-94 based on the work performed by PNL and reported in NUREG-1326 " Regulatory Analysis for the Resolution of G6neric Issue 94, Additional Low-Temperature Overpreseure Protection for Light Water Reactors," enclosed.

Low-temperature overpressure protection (LTOP) was designated as Generic Safety Issue A-26 (NUREG-0731) and resolved in 1979 by Multi 4 Plant Action B-04 (NUREG-0748).

PWR licensees were required to implement procedures to reduce the potential for overpressure events and install. equipment modifications to mitigate such events.

Current staff requirements for LTOP are in the Standard Review Plan Section 5.2.2, " Overpressure Protection," and in its attached Branch Technical Position (BTP) RSB 5-2, " Overpressure Protection of Pressurized Water Reactors While Operating at Low Temperatures" (NUREG-0800).

The administrative controls and procedures which were identified as part of i

B-04 include the following items:

i 1.

Minimize the time the reactor coolant system (RCS) is maintained in a water solid condition.

2.

Restrict the number of high pressure SI pumps operable to no more than one when the RCS is in the LTOP condition.

3.

Ensure that the steam generator to RCS temperature difference is less than 50 Deg F when a reactor coolant pump (RCP) is being started in a water solid RCS.

4 Set the PORV setpoint (if the pasticular plant relies on this component for LTOP) to a plant-specific analysis supported value, and l

have surveillance that checks the PORV actuation electronics and setpoint.

~

f B~2 Twelve PWR overpressure transients were reported during the period from 1901 to 1983, af ter completion of Generic Issue A-26.

Point Unit 4, exceeded the pressure / temperature limits of the T/S.Two of these even During this same time f rame there were 37 reported instances when at least one LTOP channel was out of service.

In 12 of these cases, both LTOP channels were inoperable.

The continuation of overpressure transient events, and the unavailability of L10P protection channels, suggested the need to re-evaluate the current overpressure protection requirements whether additional measures are warra,nted.or their implementation, to determine Major overpressurization of the Peactor Coolant System while at low temperature, if combined with a critical crack in the reactor pressure vessel welds or plate material, could result in a brittle fracture the pressure vessel.

Failure of the pressure vessel could make it impor le to provide adequate coolant to the reactor core and result in major et e damage or a core melt accident.

The safety significance of these continuing low-temperature overpressure transients was designated as Generic Issue 94, " Additional Low Temperature Overpressure Protection".

The concerns of Gl*94 are applicable to all plants regardless of the features used to mitigate a low-temperature overpressure event or of any measures to preclude events that would challenge these features or exceed the design basis for LTOP.

The implementation of the requirements for LTOP (the resolution of A-26) has been found to be essentially uniform for the Combustion Engineering (CE) and Westinghouse (W) PWRs.

With the exception of a few plants," the LTOP protection systems consist of either redundant PORVs or redundant safety relief valves (SRVs) in the residual heat removal (RHR) system and in general moet most of the requirements set forth in Branch Technical Position RSB 5-2, "Over-pressurization Protection of Pressurized Water Reactors Wnile Oper: ting at Low Temperatures."

Variability in meeting IEEE-279 requirements, equipment environmental qualification, and in meeting all the requirements of Regulatory Guide 1.26,

" Quality Group Classification and Standards for Water, Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants," exist.

As part of the NRC staff acceptance of LTOP protection system designs for the implementation of the resolution of A-26, it was concluded that the costs associated with upgrading existing systems to meet these requirements was not

• CE - San Onofre Units 2 and 3 rely on a single RHR (SOCS) SRV for LTOP.

With the SRV inoperable, depressurize and vent within eight hours.

- Maine Yankee relies on two PORVs when pressure is above 400 psig and two RHR SRVs when pressure is below 400 psig.

W - DC Cook Units 1 and 2 rely on either two PORVs or one PORV and one RHR SRV.

- Yankee Rowe relies on one PORV and two RHR SRVs.

Newer Westinghouse plants allow either two PORVs or two RHR SRVs.

ZEY f037W THE.ffKolfIe CH4CeCYMCItYtdI of YMKfACILITr ANo IM CONOlYl0AZf0 CIT.f OfdKA'f'100 YhTAEKM.KW1/CEO YG fit 0\\//M B-3 justifiable.

Further evaluations performed for GI-94 have also concluded that it is not cost beneficial to upgrade these systems to fully safety grade standards.

2.

CURRENT STANDARD TECHNICAL SPECIFICATION REQUIREMENTS The section of the Standard Technical Specifications (STS) covering the LTOP protection system is titled Overpressure Protection System Section 3.4.10.3 f or CE plants and Section 3.4.9.3 for W plants.

The LTOP system set points are established based on additional restrictions for the restart of an idle reactor coolant pump and on the number of high pressure safety injection pumps and/or coolant charging pumps allowed to be operable when LTOP is required.

These additional restrictions define the initial conditions for the plant specific transient analyses performed to establish the LTOP system set points.

The additional restrictions are provided regarding the restart of inactive reactor coolant pumps in Sections 3.4.1.3 (Hot Shutdown) and 3.4.1.4 (Cold Shutdown).

High pressure safety injection pump operability rt.strictions are provided in Section 3/4 5.3 (ECC$ Subsystems).

In addition to these administrative restrictions, the transient arialyses are based on a dual-channel system being operable to satisfy the single failure criterion of 10 CFR Part 50 Appendix A for a system which performs a safety function.

Therefore, the Overpressure Protection System T/S is consistent with Criterion 2 of the Commission's Policy Statement on Technical Specification Improvements for Nuclear Power Plants.

The T/S --- ""-- - '-- '^ adequate protection of the public health and safety in that the plant shall be operated within the bounds of the initial conditions assumed in the existing transient analyses.

The T/S also satisfied

, Criterion 3 of the Policy Statement in that the LTOP system is the primary Success path for the mitigation of low-temperature overpressure transients which present a challenge to a fission product barrier - in this case, the reactor I

pressure vessel.

1 PORVs are relied on, by most Westinghouse designed plants and about one-half of the Combustion Engineering plants, to provide LTOP protection.

In addition to l

PORVs, the RHR SRVs are also relied on to provide LTOP protection for some W plants and for the CE plants which do not have PORVs.

Newer W plants have T/S which require either two PORVs or two RHR SRVs for LTOP protection.

Administrative controls, the LCOs, concerning the RHR SRVs are identical to those for the PORVs.

The one significant difference between the PORVs and RHR SRVs is related to the surveillance requirements.

The SRVs, being passive spring loaded devices, are not functionally tested prior to entering hot shutdown.

The current STS Action requirements for the LTOP system includes a 7 day allowable outage time (A0T) to restore an inoperable LTOP channel to opertble status before other remedial measures would have to be taken.

In addition, Action d. states that the provisions of Specification 3.0.4 are not applicable.

Therefore, the plant may enter the Modes for which the Limiting Conditions for Operation (LCO) applies, during a plant shutdown or placement of the head on the vessel following refueling, when a LTOP channel is inoperable.

In this i

B-4 situation, the 7 day A0T applies for restoring the channel to operable status, before other remedial measure would have to be taken.

This is the same manner that the Action requirements apply when a LTOP channel is determined to be inoperable while the plant is in a Mode that the LTOP system is reovired to be operable.

e Based on the NRC evaluation of the LTOP system unavailability, it is concluded that additional restrictions on operation with an inoperable LTOP channel are warranted when the potential for a low temperature overpressure event is the highest, and especially when the plant is in a water solid condition.

Furthermore, it is concluded that the additional restrictions regarding the restart of inactive reactor coolant pumps and on the operability of high pressure safety injection pumps should be implemented in the T/S, as indicated in the ST$. and licensees should verify that these administrative restrictions have been implemented.

Finally, it is concluded that these additional measures will help to emphasize the importance of the LTOP system, especially while operating in a water solid condition, as the primary success path for the mitigation of overpressure transients during low-temperature operation.

3. IMPROVEMENTS IN PROTECTION SYSTEM AVAILABILITY The staff has determined that LTOP protection system unavailability is the dominant contributor to risk from low-temperature overpressure transients.

The staff has further concluded that a substantial improvenent in availability when the potential for an overpressure event is the hi'ghest, and especially during water-selid cperations, can be achieved through improved administrative restrictions on the LTOP system.

In developing the staff position on the resolution of the low-temperature overpressure protection generic issue, a number of f actors have been taken into consideration.

The staff has considered the conditions under which a low-temperature overpressure transient is most likely to occur.

While LTOP protection is required for all shutdown modes, the most vulnerable period of time was found to be Mode 5 (Cold Shutdown) with the reactor coolant temperature less than or equal to 200 Deg F, especially when water-solid, based on the detailed evaluation i

\\

of operating reactors experiences performed in support of GI-94.

LTOP transients, which have challenged the overpressure protection system, have occurred with reactor coolant temperatures in the range of 80 Deg F to 190 Deg F.

In addit ~ ion, a review of the STS for containment integrity indicates that there are no specific requirements imposed during Mode 5, when the reactor coolant temperature is below 200 Deg F.

Industry responses to Generic Letter 87-12 " Loss of RHR While RCS Partially Filled," dated July 9,1987, also indicate that containment integrity during Mode 5 is of ten relaxed to allow for testing, maintenance, and l

the repair of equipment.

In addition, the staff takes note of the fact that in all instances when pressure / temperatures limits in the T/S have been exceeded one LTOP protection channel was removed from service for maintenance related activities.

During

B-5 these events the redundant LTOP protection channel failed to mitigate the overpressure triasient as a result of a system / component failure that had not been detected.

The reported LTOP transients have occurred in Mode 5 with RCS temperatures ranging from 80 Deg F to 190 Deg F.

Since this temperature range includes Mode 6. RCS temperature less than 140 Deg F but with k-eff less than 0.95 as compared to k ef f less than 0.99 for Mode 5, the staff concludes that the additional administrative restriction for the single channel A0T is applicable to Mode 5 and Mode 6 (with the reactor pressure vessel head on).

The staff concludes that the LTOP system is important to plant safety and inoperable LTOP equipment should be restored to an operable status in a shorter period of time.

The current seven day A0T is considered to be too long under certain conoitions.

The staff has concluded that the A0T should be reduced to eight hours when operating in Mode 5 or 6, when the potential for an overpressure transient is highest.

The operating reactors experiences indicate that these events occur during planned heatup (restart of an idle reactor coolant pump) or as a result of maintenance and testing errors while in Mode 5.

The reduced A0T in Modes 5 and 6 will help to emphasize the importance of the LTOP system in mitigating overpressure transients and provide additional assurance that plant operation is consistent with the design basis transient analyses.

Based on the foregoing concerns, added assurance of L, TOP availability is to be provided by revising the cur. rent Technical Specification for Overpressure Protection to reduce the A0T for a single channel from seven days to eight hours when the plant is operating in Mode 5 or 6.

Attachment B-1 is provided for guidance for Westinghouse and CE plants.

The guidance provided is also applicable to plants which rely on both PORVs and RHR SRVs or which rely on RHR SRVs only.

Protection Technical Specification. Attachment B-2 provides the staf f bases for the O l

Licensees should also verify that the administrative controls and procedures identified in Section I have been implemented to assure that the plant is being operated within the design base.

If it is determined that the design base was developed based on restricted $1 pump operability and/or differential temperature restrictions for RCP restart ano that these restrictions have not been implemented as part of A-26, then these restrictions should be implemented now.

This is not a new requirement.

Attachment B 3 is provided.

for guidance.

l l

o B-6 Generic Issue 94 Enclosure B to Generic Letter 89-XX

_ Attachment B-1 MODIFIED STANDARD TECHNICAL SPECIFICATIONS FOR COMBUSTION ENGINEERING AND WESTINGHOUSE PLANTS

_ REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 Two power-operated relief valves (PORVs) shall be OPERABLE with a lif t setting of less than or equal to (450] psig.

APPLICABillTY:

MODE 4 when the temperature of any RCS cold leg is less than and the RCS is not vented through aor equal to (275)

'F, MODE 5, and MODE 6 when the head is on the reactor vessel square inch or larger vent.

ACTION:

With one PORV inoperable in MODE 4 restore 'the inoperable PORV to a.

OPERABLE status within 7 days or depressurize and vent the RCS i

through at least a square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, b.

With one PORV inoperable in MODES 5 or 6 restore the inoperable PORV to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or depressurize and vent the RCS through at least a hours.

square inch vent within the next 8 With both PORVs inoperable, depressurize and vent the RCS through at c.

least a square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, d.

With the RCS vented per ACTIONS a, b, or c verify the vent pathway at least once per 31 days when the pathway is provided by a valve (s) i which is locked, sealed, or otherwise secured in the open position;.

otherwise verify the vent pathway every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

1 In the event either the PORVs or the RCS vent (s) are used to mitigate e.

i an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within t

30 days.

transient, the effect of the PORVs or RCS vent (s) on the tra and any corrective action necessary to prevent recurrence, f.

The provisions of Specification 3.0.4 are not applicable.

-.m

B-7 Generic Issue 94

$URVEILLANCE REQUIREMENTS 4.4.9.3 Each PORV thall be demonstrated OPERABLE by:

Performance of an ANALOG CHANNEL OPERATIONAL TEST, but excluding a.

valve operation, at least once per 31 days.

6. -

Performance of 4 CHANNEL CALIBRATION at.least once per 18 months; and Verifying the PORV isolation valve is open at least once per'72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

c.

I G

I I

b B-8 Generic Issue 94 Enclosure B to Generic Letter 89-XX Attachment B-2 3/4.4.9.3 OVERPRESSURE PROTECTION SYSTEM Bases of the Limitino Condition _ for Operation and Surveillance Requirements:

The OPERABILITY of the PORVs is determined on the basis of their being capable of performing the function to mitigate an overpressure event during low-temperature operation.

The Modified Standard Technical Specification (STS) requirements include the following changes from prior STS guidance:

1.

The depressurizing and venting of the RCS is not classified as an overpressure protection system.

However, the APPLICABILITY of the LCO excludes Modo 6 when the RCS is adequately vented.

This avoids any possible question on Specification 3.0.4 being applied to preclude placement of the head on the vessel if any part of the LCO is not met when the RCS is vented.

2.

The APPLICABILITY for MODE 6 is clarified as "when the head is on the reactor vessel" rather than as " MODE 6 with the reactor vessel head on."

3.

ACTION a, s revised to clarify that it is only. applicable in MODE 4.

4 ACTION b. was auded to reduce the allowed outage time for an inoperable PORV to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in MODES 5 or 6.-

Because this LCO does not apply under certain conditions specified under the APPLICABILITY for this l

specification, the ACTION statements likewise do not apply under those conditions.

ACTIONS a. and b. do not repeat those qualifying conditions that apply for these MODES since the ACTIONS only apply when the unit is in those conditions.

5.

ACTION d. includes the requirements to verify that ACTIONS a, b, or c continue to be met on an ongoing basis when the unit would be in MODES 4, 5 or 6.

6.

The Surveillance Requirements were simplified by removing requirements that exist due to the general requirements applicable to all surveillance requirements as specified in Section 4.0 of the TS.

7.

Surveillance Requirement 4.4.9.3.2 was removed since it is addressed by ACTION d.

For plants with existing TS for PORVs used for LTOP, the only required change is that indicated to restrict the applicability of ACTION a to Mode 4 and for incorporating ACTION b.

Any other changes that are proposed consistent with s

m., -

w

1 B-9 Generic Issue 94 the above guidance are voluntary.

For a plant without existing TS for PORVs that are used for LTOP, a TS should be proposed that conform to the above guidance.

Because some plants use residual heat removal (RHR) suction relief valves for LTOP, either in addition to or in lieu of PORVs, similar requirerents are included in TS as noted above for PORVs.

a) and b) are required, as noted above, for these plants.The same changes in ACTION

~

Likewise, any plant without existing TS for RHR suction relief valves that are used for LTOP, should propose TS that are consistent with the above guidance.-

When only RHR suction relief valves are used for LTOP, the Surveillance Requirements would state "No additional requirements other than those required by Specification 4.0.5."

J I

v

1 B-10 Generic Issue 94 Enciesure B to Generic Letter 89-XX Attachment B-3 STANDARD TECHNICAL SPECIFICATIONS GUIDANCE l

FOR COMBUSTION ENGINEERING AND WESTINGHOUSE PLANTS Operational Limitations Consistent With The Design Basis Assumptions For The Low-temperature Overpressure Protection (LTOP) System The TS requirements for LTOP typically apply in Mode 4 when the temperature of any cold leg is below 275 'F, Mode 5, and Mode 6 when the head is on the reactor vessel.

During these conditions, one train (or channel) of the LTOP system is capable of mitigating an LTOP event that is bounded by the largest mass addition to the RCS or of the largest increase in RCS temperature that can The largest mass addition to the RCS is limited based upon the assamp-occur.

tion that no more than a fixed number of pumps are capable of providing makeup or injection into the RCS.

Hence, this is a matter which is important to safety that pumps in excess of this design basis assumption for LTOP not be capable of providing makeup or injection to the RCS.

The capability for makeup and injection to the RCS is also a safety concern for normal makeup to the reactor coolant system for reactivity control as well as for events would could result in a loss of coolant from the RCS.

The former are covered by Technical Specifications (TS) under Reactivity Control i

Systems, Charging Pump - Shutdown (Modes 5 & 6); Charging Pumps - Operating (Modes 1 through 4); and Flow Paths - Operating (Modes 1 through 4).

The latter is covered by TS under Emergency Core Cooling Systems, ECCS Subsystems -

Tcold less Than 350'F (Mode 4).

The manner in which restrictions,-consistent with the design basis assumptions of the LTOP system, have been incorporated in TS, that_ require the operability l

of makeup or injection pumps, has varied depending upon plant-specific l

considerations for the LTOP design and plant-specific designs for the use of pumps for makeup and ECCS functions.

A common method has been the use of footnotes to the pump operability requirements to note that:

A maximum of one Safety injection [and/or] one charging pump shall be OPERABLE when the temperature of one or more of the RCS cold legs is less than 275'F.

This footnote is used for each specification that requires the operability of a safety injection and/or charging pump in Modes 4 or 5.

The Surveillance Requirements typically include the following:

All Safety injection [and/or] charging pumps, except the above required OPERABLEpump[s),shallbedemonstratedtobeinoperablebyverifying e

n

4 B-11 Generic Issue 94 that the motor circuit breakers are secured in the open position at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperature of one or more of the RCS cold legs is less than or equal to 275'F.

Generally, it is preferable to include requirements for implementing the intent of an LCO as part of an LCO rather than to only define requirements, such as securing motor circuit breakers in the open position, in a Surveillance Requirement.

Furthermore, the requirements for operable pumps could be stated in terms of requiring one pump to be operable rather in terms of "at least one pump shall be operable" and then including a footnote requiring that, in fact, no more than one pump shall be operable.

The preferred alternative would be an LCO which stated:

One Safaty Injection (and/or) charging pump shall be operable and all other Safety injection (and/or) charging pumps shall be secured with their motor circuit breakers in the open position.

The form of the above requirements for any given specification would be dependent upon which pumps are addressed by that specification, e.g. charging or injection pumps or both.

The surveillance requirements would be similar to that noted above with the following substitution:

...except the above required OPERABLE pump (s), shall be demonstrated to be Occured by verifying that the motor circuit breakers are in the open position....

i Changes to plant TS should proposed to incorporate one of the above methods, for ensuring that pumps are not capable of initiating a mass addition to the RCS that exceeds the design basis assumptions for the LTOP system, for plants that do not currently include such requirements.

The largest temperature increase in the RCS that could result in a challenge to the LTOP system is dependent upon the differential temperature between the L

RCS and the secondary system when starting a reactor coolant pump.

Hence, this is also a matter which is important to safety when reactor coolant pumps are started and the resulting increase in RCS temperature is in excess of the~

design basis assumption for the LTOP system to mitigate the resulting increase in RCS pressure.

The manner in which this design basis assumption of the LTOP system is reflected in TS has been the use of a footnote to the reactor coolant 1

pump operability requirements to note that:

A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 275'F unless the secondary water temperature of each steam generator is less than

• F above each of the RCS cold leg temperatures.

c-B-12 Generic Issue 94 The above footnote has been included in the TS for residual heat removal under title of the Reactor Coolant System Hot Shutdown.

Changes to plant TS should proposed to incorporate the above method, to ensure that exceeds the design basis assumptions for the LTOP sys do not currently have this requirement, d

e e

i

)

i As

E

+ys teo u

l=

?,

l' UNITED STATES NUCLEAR REGULATORY COMMISSION s,

WASHINGTON. D. C. 20666

/

AUG t 21989 MEMORANDUM FOR:

Edward L. Jordan, Chairman Committee to Review Generic Requirements THROUGH:

R. Wayne Houston, Director je Division of Safety Issue Resolution,v /, '

Office of ts.: lear Regulatory Research FROM:

Robert L. Baer, Chief Engineering issues Branch Division of Safety Issue Resolution Office of Nuclear Regulatory Research

SUBJECT:

RISK / BENEFIT ANALYSIS OF KEEPING POWER AVAILABLE TO PORV BLOCK VALVE

Reference:

Shutdown Decay Heat Removal Anal Plant Case Studies and Special Issues,ysis:

NUREG/CR-5230, April 1989 Currently, there is considerable variation in the technical specifications (TS) for PWRs regarding PORVs and their block valves. 'As part of the resolution of Generic !ssue 70 (GI-70), we are proposing that the TS be made uniform for all PWRs having PORVs.

One element of this would be to still require block valves to be closed in the event of a leaking PORY but require that power not be removed from the block valve.

The advantage, as pointed out in our proposed resolution package, is that feed and bleed operation would be more reliably achievable with this mode of operation.

The reference (NUREG/CR-5230) reports that feed and bleed operation can provide a si core melt probability if initiated in a timely manner.gnificant reduction in However, keeping power available to the block valves in the event of leaking PORVs does increase the probability of inadvertent opening of the block valves and could conceivably result in a small LOCA.

This memorandum reports on our evaluation of the competing risks and benefits described above as requested by the CRGR during the August 9,1989 meeting on GI-70. In summary, the benefit in keeping power available to the block valves is much greater than the increased risk, a net reduction in core damage frequency (CDF) of IE 6 to 3E-5/ reactor year.

The risk of LOCA because of inadvertent openin a leaking PORV was estimated to be quite low. g of the block valve that isolates The event tree for this scenario is shown in Figure 1.

Both the best estimate and a conservative estimate of the probability of S small LOCA occurring are shown in Figure 1.

(Conservative estimate values ar..,hown in parenthesis.)

To obtain the core melt probability, the values in Figure 1 were multiplied by the conditional probability of CDF, given a small LOCA.

Using 3E-3 for a typical PWR, the best estimate of CDF is to Enclosure 2 l

j 2

AUG 2 2 IgEg about 3E-9 per reactor-year.

Using a conservative conditional probability of 8E-3 and the conservative probability of the occurrence of a small LOCA from Figure 1 results in a CDF of 2E-7 per reactor-year.

Feed and bleed operation is needed for events in which heat removal by the steam generator has been lost (e.g., loss of main feedwater and failure of all auxiliary feedwater).

The time available to initiate feed and bleed is dependent on the type of NSSS and the event scenario.

This is shown in Table 1 which was developed from information presented in NUREG/CR-5230.

To be effective, feed and bleed should be initiated by the time the shell side of the steam generator boils dry and must be initiated in W and CE plants with PORVs before the RCS becomes saturated and the RCS pressure increases to HPI shutoff head.

Regarding the 10 to 15 minute time periods for W and CE plants.

NUREG/CR-5230 states, "It would appear that this puts a considerable pressure upon the operators. A decision to open the primary must be made before there is really any time to establish reasons for other equipment not functioning."

Although all B&W steam generators would boil dry in 5-8 minutes, the HP! pump shutoff head is above the safety valve setpoint in all B&W plants except Davis Besse.

The probability of achieving feed and bleed operation was evaluated as a function of "available time" for (1) the case where power was maintained to the block valve and (2) the case where power was " racked-out" at the motor control The "available time" represents the period aft'er a decision was made to center.

initiate feed and bleed and was varied from 6 minutes to 20 minutes.

The event tree used is shown in Figure 2 with the results for an "available time" of 12.5 minutes shown for Case 1 and, in parentheses Case 2.

Results for other times are presented in Table 2.

Information from the referenced report was used to obtain an estimate of the reduction in core melt frequency associated with feed and bleed operation.

That report analyzed four plants and showed that successful operation of feed and bleed reduced CDF frequency from 1.15E-3 to 4.8E-5.

For the purposes of this analysis 1E-4 was used.

This value was then reduced to 4E-5, because plants operate with block valves closed about 50% of the time, and it was assumed that successful feed and bleed operation would be reduced by another 20% because of equipment failures.

Multiplying the right hand column of Table 2 by 4E-5 resulted in an estimated reduction in CDF probability in the range of 1.1E-6 to 2.7E-5 if power is maintained to the block valve after it has been closed.

The low end of this range exceeds by a factor of 5 the conservative estimate of the increased CDF probability due to a possible LOCA, because power was maintained to the block valve.

i However, the reduction in core damage frequency of 4E-5/ reactor year due to i

feed and bleed is based on the assumption that operator will rarely fail to i

decide to initiate feed and bleed; i.e., a human error probability (HEP) of 3E-3 was used in the referenced report.

However, operators have failed to decide to initiate feed and bleed in at least half of the actual cases.

If the l

AUG t 21;;g HEP is 0.5, then the reduction in CDF is 2.4E-5 instead of 4E-5. -Therefore, if operators are hesitant to use feed and bleed the condition of the block valve is less important, and leaving power on the valve reduces C0F by 0.7E-6 to 1.6E-5/ reactor year.

However, the conclusion regarding relative risk is unchanged.

• w d'x'

'1 e

Robert L. Baer. Chief Engineering Issues Branch Division of Safety Issue Resolution Office of Nuclear Regulatory Research i

4

i

)

Table 1 Time to Boil Steam Generator Dry (Minutes)

With Nominal Scram at

NSSS, SG Inventory Low level Setpoint B&W 8

5 W&CE 30 to 40, but 10-15 minimum as much as 46 for one plant i

Table 2 Estimated Probability of Initiating Feed and Bleed Time Available Estimated Probability After Decision With Power With Power Change Made to Achieve Maintained

" Racked-Out in 6

Feed and Bleed to Block Valve Initially Probability 20

.998

.970

.028 15

.998

.856

.142 12.5

.998

.799

.199 10

.994

.692

.302 8

.988

.624

.364 l

6

.985

.315

.670 l

l t

Figure 1 Inadvertent Opening of Block Valve 4

Leaking PORV.

Block PORV Block Conseque~nces Block Valve Valve fails Valve Closed Opened Open Reclosed (See Note 1) (See Notes (See Note 4) (See Note 5) 2 & 3)

No leakage

.013

(,06) i

.0016

.045 9.4E-7 Small LOCA t.01)

(.2)

(2.7E-5)

Leaking PORV, limits set by existing Tech Specs No leakage Notes:

1.

Assumed initial condition with power maintained to block valve 2.

Best estimate probability based on the fact that there has been about 600 years of operation with a block valve closed for plants that do not have a TS requirement to " rack-out" power.

There are no known incidents of bitck valve being inadvertently opened.

3 Probbility estimates exclude sudden structural failure of seat / disk since this is a very low probability event, and the probability is not affected if power is maintained or " racked-out."

4.

Probability estimates are the sum of:

a.

Sudden opening of PORV when block valve is opened.

The range was judged to be.01 to.001, with a best estimate of.003.

b.

Inadvertent opening of block valve not noticed for a period of time, i

transient occurs that opens PORV, and PORY fails to close (.01 best i

estimate and.05 conservative estimate, both based on judgement).

5.

Best estimate based on IEB 85-03 results, INEL tests that showed that flow was reduced greatly even when valve did not fully close, and the fact that a very defective valve operator would not be able _to open the block valve initially. Conservative estimate based on BNL study.

Figure 2 Probability of Initiating feed & Bleed Decision Restore Block Other Operator Success'ul f

to initiate Power to Valse Actions initiation feed and Block Valve Opened in Performed of Feed & Bleed Bleed Specified in Specified Time Time

.999

.998

(.945)

(.799)

(

.999

(.995) 1.0

(.85)

NO No No Notes:

1.

Probabilities shown are for a total available time of 12.5 minutes.

2.

Probabilities for case with power available to the block valve are above the horizontal lines in the event tree.

Probabilities for the case with power " racked-out" to block valve are shown in parenthesis below the horizontal lines, i

l I

I

, to the Minutes of CRGR Meeting No. 168 Briefing on Treatment of External Events in the ITE August 23, 1989 i

TOPIC D. Ross (RES) and L. Shao (RES) briefed the Committee on issues to be considered in developing guidance for the treatment of external events in the IPE, and milestones of the projected schedule for that guidance development

(

effort.

Briefing slides used in the presentations to the Committee are enclosed (Attachments 1 and 2).

BACKGROUND The following background documents were provided to the Committee in connection with this IPEEE briefing:

1.

Memorandum dated June 17, 1989, D. Ross to E. Jordan et al., and attachment (dated June 15, 1989) entitled "On External Events."

2.

Draft letter dated July 18, 1989, D. Ross (NRC) to R. Chung (National Research Council).

CONCLUSIONS / RECOMMENDATIONS i

No recommendations were made by the Committee to the EDO as a result of this briefing.

The Committee requested to be kept informed regarding the status of development of IPEEE guidance.

i i

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EXTERNAL EVENTS CRGR l

Aug 23,1989 bN Ei a

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.a w

u--,-..----

.w,....

s STATEMENT OF PROBLEM

• Per IPE Letter, all licensees must examine for particular vulnerabilities to core damage j

or unusual containment performance

• Many risk studies show seismic or fire can be l

important or dominate (see Table).

• Guidance for external events under development

• Lots of uncertainty on proper hazard curve

• Methods for assessing seismic margin essentially complete i

but will need definitive guidance document.

I

• Some seismic mods needed, but not overwhelming.

O L

r m

i l

SEISMIC I

• Multiyear effort by NRC and EPRI have produced estimates of seismic hazard at many sites.

• Estimates of hazard vary widely in the two studies.

• Consequent estimates of c.d.f. thus vary (see table)

• Who is right? Is that a fair question?

• Does NRC need c.d.f., or can we do IPEEE without one?

• Present belief is either seismic margin or PRA.

l 4

r A

i TABLE 3.1

• • W.A. = WOT AVAILABLE ** PIANTS WITN PRA'S AVAf TABLE TO NRC STAFF ANALYST /

PLANT PRA INTERNAL-SEISMIC FidDOD FIRE 10FAL PLANT PROGRAN TYPE LEVEL CNF(E-4) CNF(E-4) CNF(E-4) CNF(E-4)

CNF(E-4)

ANO-1 IREP B&W 2

0.42 W.A.

N.A.

N.A.

N.A..

ANO-1 TAP A-45 B&W 3

0.88 0.73 0.072 0.058 1.79 BIG ROCK POINT INDUSTRY BWR1 3

9.75 N.A.

N.A.

2.3 N.A.

BROWNS FERRY-1 IREP BWR4 NK1 1

2.0 N.A.

N.A.

N.A.

N.A.

BROWNS FERRY-1 INDUSTRY BWR4 NK1 3

15.0 1.4 0.012 1.5 17.9 BRUNSWICK 1 INDUSTRY BWR4 NK1 1

0.25 N.A.

N.A.

W.A.

M.A.

BRUNSWICK 2 INDUSTRY BWR4 NK1 1

0.21 N.A.

N.A.

N.A.

W.A.

CALVERT CLIFPS IREP CE 1

1.3 N.A.

W.A.

N.A.

N.A.

CALVERT CLIFFS RSSMAP CE 1

20 N.A.

N.A.

W.A.

M.A.

CESSAR INDUSTRY CE SYS89&

1 0.81 N.A.

W.A.

W.A.

N.A.

COOPER TAP A-45 BWR4 NK1 3

2.9 0.81 0.50 0.11 4.4 CRYSTAL RITER-3 IREP B&W 2

4 N.A.

N.A.

N.A.

M.A.

CRYSTAL *tIVER-3 INDUSTRY B&W 1

1.1 N.A.

N.A.

W.A.

. N.A.

GESSAR II INDUSTRY BWR 3

.05 N.A.

N.A.

W.A.

N.A.

GRAND GULF-1 IDCOR BWR6 NK3 3

.08 N.A.

W.A.

N.A.

M.A.

GRAND GULF-1 RSSNAP BWR6 NK3 1

.4 N.A.

N.A.

N.A.

N.A.

GRAND GULF-1 NUREG-1150 BWR6 NF.3 3

.3 N.A.

N.A.

N.A.

N.A.

INDIAN POINT-2 INDUSTRY W4 3

0.9 1.4 N.A.

2.0 4.7 INDIAN POINT-3 INDUSTRY M4 3

1.3 0.03 N.A.

0.63 1.97 IASALLE-2 RMIEP BWR5 NK2 3

N.A.

N.S.

N.A.

N.A.

N.A.

LINERICK-1 INDUSTRY BWR4 Nd2 3

0.15 N.A.

N.A.

N.A.

W.A.

NILLSPDNE-1 IREP BWR3 NK1 1

3 N.A.

N.A.

M.A.

N.A.

NILLS 1DNE-3 INDUSTRY W4 7

9.46 N.A.

W.A.

W.A.

N.A.

OCONEE-3 RSSNAP B&W 2

0.8 N.A.

N.A.

N.A.

N.A.

OCONEE-3 EPRI/MSAC B&W 3

1.2 N.A.

N.A.

W.A.

N.A.

PEACH Boff 0N RSS BWR4 NK1 3

.3 N.A.

W.A.

N.A.

N.A.

PEACM BOTTON IDCOR BWR4 NK1 3

0.36 N.A.

N.A.

N.A.

W.A.

PEACH BUPPDN NUREG-1150 BWR4 NK1 3

.09 N.A.

N.A.

W.A.

N.A.

POINT BEACN-1 TAP A-45 W2 3

1.4

.61

.77 0.33 3.1 QUAD CITIES-1 TAP A-45 BWR3 NK1 3

.0.99

.83 0.001 0.13 2.0 2.3 0.28

.04 0.25 2.9 SEABROOK INDUSTRY W4 3

SEQUOYAN IDCOR W4 IC 3

.9 N.A.

N.A.

N.A.

N.A.

SF400YAN RD%P W4 IC 1

.6 N.A.

N.A.

W.A.

N.A.

SE(UOYAN WJ:G-1150 W4 IC 3

1.0 N.A.

N.A.

N.A.

W.A.

SNORENAN A CfSTRY BWR4 NK2 3

0.55 N.A.

N.A.

N.A.

N.A.

SP-90 ICLUSTRY APWR 1

0.06 N.A.

W.A.

N.A.

N.A.

ST. LUCIE-1 TAP A-45 CE 3

.14

.13 N.A.

.44

.74 SURRY NUREG-1150 N3 SA 3

.26 N.A.

N.A.

W.A.

N.A.

RSS W3SA 3

0.44 W.A.

N.A.

W.A.

N.A.

TNI-1 INDUSTRY B&W 7

4.4

.027

.075 0.9 5.5 SURRY 1URKEY POINT-1 TAP A-45 W3 3

.71 0.073

.46

.75 2.36 YANKEE ROWE INDUSTRY M4 3

.02 N.A.

N.A.

N.A.

N.A.

ZION IDCOR W4 3

1 N.A.

N.A.

W.A.

N.A.

ZION INDUSTRY W4 3

9.43 0.96 N.A.

W.A.

W.A.

ZION NUREG-1150 W4 3

1.5 N.A.

W.A.

W.A.

N.A.

a

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1 I

)

f IPRI HAZARD CURyts l

10

r p

i g

j PEACH Bo7?og

/

8 10"

-- - = MEAN HAZARD

.\\.

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h

s,s '.,

e.;, s.

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\\ s., ~.s.

4

\\,g' s Mean Curve 10's O'~, N s'-

,, '. s, s *..

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,,,N...,.....-

N r

s 3;

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=e ACCELERATION CM/SEC*"2 Figure 3.)

.fj Mu. d80. and ames pesa ground asseAoresten I

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t.W.S StitWit 14684A9 WAR &ttttilafl8N L0rta te6hifWDt Of INit0A&fl0N IS 6.0 LLNL i

1 06Atate SWWtl UllNG &LL (IPttfl l

10 1

D4ttf.64Riftt6 fit.64tehETRic at i

10 k

p

( \\ '.0 <

a I

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'.3 i

• /

.a I10

=d l

t

( -. 3E(-5) l

-6 g to I

4 10 1

7 1

to

.a n

n e

e a

e Tw AcettenAfl0N CW/StC'2 l

PEACH BOTTN Figure 3.2

-.a-.,.

-..,. ~,. - -. - - -,..,,.,,, - -.. -..

~.....,--..,. -.

-..~. _..,.,--....,,.,.~.

w

[PRI HAZAtD CURVES t

gg6 -

I gr, guggy

---

19 WZAN 10*

b G

s Iif g

%g' P

s\\.

r 1[ [

gg 8

AM N0 90 88 90 90 100 WO 9001000 MMM W-5 Figure 3.3.

Annual probability of exceedence of peak ground acceleration.

i e

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s; t.U.S StlSwit H&tARD Ct44t&CTttitatl0N I

LOWtt teAGNIflJDC Of $NTCORA100N I$ $.0 LLNL l

MA1&tD CUWtl Ullt4 ALL (IPERTS e

1 to

$=4t$1 A=&hlTIR41tC,64tchCTRIC d

10 I

W E

3 l

l E 10 y

5

. {10 B

,.=

.s a

g to f.

to 7

a n

e

=

to g

l acctLtnatioN CW5tC2 t-SURRY Figure 3,4 4

d

_, _.,,.... - +,.. -. -.,

.,.~..-.__,_-..,,...%..

,-,r,..-..

,,m.,..

.--r...

I l

Table 3.6 y

Core Dama9e Frequencies l

I

{-

g g

Median.

95th Mean LLNL 3.9tt-7 1.48E-5 4.38E-4 1.16E 4 i

k EPR) 3.00E.7 6.12E 6 1.03E-4 2.50E 5'

{

Fire 2.2E-6 8.32E-6 3.08E-5 1.13E-5 Isternal 6.80E 6 2.30E 5 1.30E-4

'4.10E-5 Peach Bottom LLNL 5.33E-8 4.41E-6 2.72E-4 7.66E-5 i

EPR1 2.30E-8 7.07E-7 1.27E-5 3.09E-6 Fire 1.09t 6 1.16E-5 6.37E-5 1.96E-5 Interna) 3.50E-7 1.90[ 6 1.30E-5 4.50E 6 e

s u

\\

L NUREG -1150 ON SEISMIC AND FIRE

• See Plant Damage State Matrices 4

i

• Appreciable numbers for both seismic and fire

• Appreciable effect on early cont. failure, seismic, Surry

• Peer review for seismic portion not yet defined.

1 i

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?

M -

~

.a

.<et.

?g$

u" e.

g g?

-.u G

h. O c Og I

O'

.-E-h'

= 115 s.

a p' 8 E -

~8 x

w-6e I'.v I-ehw' I R>

cu m$QV CC Da o

b w

o 8

+

=g w

En-e U

f 3

a O

O

~

me o

o r

'W c===i:n l

a

. a.

gg CO 9

m M

Ws W

w N-g e,

b 0

N bW W

=

C o

E i

w 15 O

gg C

w i

t O

02E Eb Eb Es kb 8

[

p m

B g

b E k" k

k t

n.

i u

.O e

e _e 2

- u g3.

m d

4 si o

0U eB e *E e 'I" e E 2

U <t' c., e

>e o

h' i

-.,-,..-...,--e.,

-,,=mm e-

I l

Probability Conditional on Early Containment Follure M - mean m - median

-l l

=

=

n I

g IE0 -

tE-t j

)

-l

.)

1 l

' l IE-2 IE-3

[

IE-4 t!'

i i

i I

(IE-4

)

i

+

I' i

SelSmlC j

Plant Demoge Flre State (1.9eE-6)

- (7.62E-8)

(Core Damage Freetseny) l 3.6 Conditional Probabl19ty Distributions I

rfgure For Early Contalement Fallere (Peach Bottom) t t

l i

a.

I k

l

.i Plant Domoge State Collapsed t u. o o c o,. o o. o o. re. a.. a w )

Accident Progression Btns Fire Selsmic (l.coE-6)

(19.20E-6)

I V8.olpha.

006

.006 i

early CF V8) 200 pet.

013

.006

{

j early CF -

.i i

V8 ( 200 pet.

.082 early CF 3

V8. BMT. late CL

.20

.26 1

1 l

Bypass

.001 f

V8. no CF

.60

.436 l

No V8 189 I

i Key: BMT - Bosemat Mell-Through CF - Containment Fatture CL - ContotnmenI Leot i

V8-- Vesset Brooch

~

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Figure 3.7 Conditional Proba 111ty)of Accident Progression Bins Servy i

~

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Probability Conditional on Early Contolnment Failure t

u - m.on m - median i

t I

IE0 -

l

,t.

[

IE-1 w-et -

~

IE-2 l

1

+

.s-i IE-3

>=

IE-4 1

i

.(tE-4 i

Plant OofsoGe Fire Selsmic Stole (1.09E-5)

(19.20E-6) i (core Domage Frequeny).

3.8 Conditional Probability Distribptions Figure For Early Containment'Fallere <5erry) j

.j l

,j

q i

1 i

1.0E a 03

,a 1

5 i

E

~

~~

s 1.0E 04 I

e a

a i

I 1.0E.06 r

g g.

l

• 0E 06 c

1.0E 07 INTERNAL SEISWIC SEISMIC FIRE uvERMoRE RPRI i

6 Mean O Median Fiture

' burry external events, sore damage frequency ranges (6th and 96th percentiles)

~

1.0E 03 g

h 1.0E 04 t

a E

o 0

1.0E 05 t.

~

~

~~

^

f E

1.0E 06 [

~~

l V

E 1.0E 07 t

INTERNAL SEISMIC SEISMIC FIRE uv8AMoRE EPRI O Mean O Median Figure

• hench Bottom external events, core damage frequency ranges (6th and 95th percentiles) e

0.3 a-0.28 e

4 0.,

I O.15

.2 a

T 0.1 1

4 4*

0.0s 7

10E-08 10E 07 1.0E 1.0E-08 10544 10E-03 10E-02 CORE DAMAGE FREQUENCY

~ Stl8Mio. LIVERMORE

+- SilSMIC. EPRI '"*= F IR E

~ Figure 3.9 Sutry enternal events sore dainage frequency distributions 0.2 s e

I 0.2 i

{

0.15 i.

0.1 e.

t; 0.05

_ = _...

10E-08 10E-07 1.0E-06 10E-06 1.0E-04 10E-03 10E-02 CORE DAMAGE FREQUENCY

~ 8tiSMIC. Livsmuont

-+- Sil8Mic. RPRI **- F IR E F6gure

• heech Bottom enternal events core dama0e frequency distributions

7 SEISMIC MARGIN i

• L Shao will discuss methodology l

• Margins methods a product of RES over past few years

• Test cases include Maine Yankee,-and Hatch to some level.

• Does not make use of extreme end of hazard curve 4

i ia l

l

.i

+

-1 FIRE

• L Shao to discuss IPEEE solution

• Even Appendix R plants have comparable c.d.f

• Expectation is some mods needed, but no change to App R l

l 1

l

.{

b i

i i

4 i

2

11.

' i POLICY

~

• ls IPEEE question ~of sdequa'cy or enhuhc6nierit?

~ ~ ~ ~

• Do we need core damage frequency?

l

• Are methods mature enough to pulse the industry?

• ls more research needed?

• What is the role of the safety goal?

1 l

• Do we need to modify any regulations?

I 1

a l-S g

l 1

i s

.. ~- -

SG

(

c i

i TREATMENT OF EXTERNAL EVENTS IN THE SEVERE ACCIDENT PROGRAM LAWRENCE C. SHA0 AUGUST 23, 1989

_1

.. -. Attachment _2_ to Enclosure 3.'. _.,

(

l SEVERE ACCIDEllT POLICY STATEMENTS NO DIFFERENTIATION BETWEEN INTERNAL AND EXTERNAL EVENT l

INDIV1 DUAL PLANT EXAMINATION GENERIC LETTER MENTIONS EXTERNAL EVENTS 2-e

NP.C EXTERNAL EVENTS STEERING GROUP (EESG)

MISSION MAKE RECOMMENDAT10tlS TO NRC SENIOR MANAGEMENT REGARDING:

ROLE OF EXTERNAL EVENTS WITHIN NRC SEVERE ACCIDENTS POLICY GUIDANCES FOR IMPLEMENTATION OF EXTERNAL EVENTS INDIVIDUAL PLANT EXAMINATION (IPEEEs)

INTEGRATI0tl 0F ALL NRC EXTERNAL EVENT PROGRAMS NEEDED RESEARCH OR TECHNICAL ASSISTANCE

l i-KEY EXTERNAL EVENTS I

i i

EARTHOUAKES j

INTERNAL FIRES EXTERNAL FLOODS i

WIND AND TORNADOS TRANSPORTATION ACCIDENTS AND OTHERS i

i J

4 e

L

l NRC EXTERNAL EVENTS STEERiflG GROUP MEMBERSHIP I

^

CHAIRMAN:

L. C. SHA0 NRR MEMBERS:

G. ARLOTTO RES T. NOVAK AE0D.

R. W. HOUSTON RES EXECUTIVE:

G BAGCHI NRR SECRETARY:

h SEISMIC SUBCOMMITTEE:

CO-CHAIRMEN:

L..REITER '

NRR I

A. MURPHY RES FIRE SUBCOMMITTEE:

CHAIRMAN:

C MCCRACKEN NRR i

HIGH WIND, FLOOD AND c

OTHERS SUBCOMMITTEE:

CO-CHAIRMEN:

D. JENG NRR W. BECKNER RES

-S-f.

i

s NUCLEAR ~lNDUSTRY'S COUNTERPART ORGANIZATION

-i NUCLEAR UTILITY MANAGEMENT AND RESOURCES COUNCIL (NUMARC)

SEISMIC ISSUES SEVERE ACCIDENT WORKING GROUP WORKING GROUP

. CHAIRMAN:

W. LINDBLAD CHAIRMAN:

CORDELL REED 1

RESPONSIBLE FOR RESOLUTION OF RESPONSIBLE FOR: RESOLUTION ALL SEISMIC ISSUES OF OTHER EXTERNAL EVENTS ISSUES AND. ACCIDENT MANAGEMENT l

1 i

6-t

.-..--6

..-w,

]

l EXTERNAL EVENTS C0HSIDERAT10flS MANY POSSIBLE SOURCES OF HAZARDS-LARGE UNCERTAINTIES ON FREQUENCY OF INITIATING EVENTS 4

PLANTS. DESIGNED TO VARIOUS CRITERIA-5,,, s..

+, s3

. EXTENT OF PROTECTION WITH REGARD TO EXTtsNAL EVENTS i

BEYOND DESIGN BASES ARE UNKNOWN FOR MANY PLANTS (MAY-NOT BE CONSISTENT) t

~

PRAs INDICATE RISKS DUE TO CERTAIN EXTERNAL. EVENTS CAN BE HIGH NEED APPROACHES FOR. EVALUATING VARIOUS EXTERNAL EVENTS BEYOND DESIGN BASES-EXTERNAL EVENTS PROGRAMS NEED TO BE INTEGRATED 1

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SEISMIC DESIGH VINTAGE PRE 1962 PLANTS:

NO SPECIFIC CONSIDERATION FOR SEISMIC

-DESIGN, 1963 TO 1969 PLANTS:

USED UNIFORM BUILDING CODE OR EQUIVALENT' J

STATIC METHOD OF ANALYSIS, 1970 T0 1975 PLANTS:

USED HOUSNER'S RESPONSE SPECTRUM CURVE'S FOR SEISMIC INPUT, CRITERIA USED FOR EQUIPMENT QUALIFICATION VARIES 1975 TO CURRENT:

R G. 1.60 SPECTRUMLOR EQUIVALENT USED FOR SEISMIC INPUT, EQUIPMENT QUALIFICATION IN ACCORDANCE WITH IEEE:344-1975, SEISMIC REEVALUATION EFFORT:

10 EARLIEST PLANTS UNDER SYSTEMATIC EVALUATION PROGRAM (BIG ROCK POINT, DRESDEN 1 a 2, HADDAM NECK, GINNA, LA CROSSE, PALISADES, 0YSTER CREEK, SAN ONOFRE 1, MILLSTONE 1, YANKEE R0WE) WERE EVALUATED TO MEET A SET OF SEISMIC CRITERIA (RELAXED FROM SRP CRITERIA)'WHICH WILL' ENSURE SAFE SHUTDOWN CAPABILITY AND PRESSURE-BOUNDAR INTEGRITY i

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i SEISMIC IPEEE

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AVAILABLE METHODOLOGY SEISMIC PRA SEISMIC RISKS M.s

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SEISMIC MARGINS PLANT VULNERABILITIES T0-SEISMIC EVENTS-9 e

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SElSMIC MARGIN METHODOLOGY o

ASSESS THE INHERENT CAPABILITIES OF NUCLEAR PLANTS TO WITHSTAND EARTHQUAKES AB0VE THE DESIGN LEVEL o

EFFECTIVE AND EFFICIENT MEANS.TO'lDENTIFY SEISMIC VULNERABILITIES o

VULNERABILITIES-DEFINED.lN HCLPF TERM o' NO SEISMIC HAZARD CURVES USED' o

NO RISK NUMBERS

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SYSTEMS AND COMPONENTS REVIEWED BASED ON PAST.

PRA EXPERIENCE o

RIG 0ROUS PLANT WALKDOWNS o

KEY' ISSUE:

REVIEW LEVEL EARTHOUAKES i

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SEISMIC DESIGN MARGIN TRIAL PLANT REVIEWS INITIAL PLANT FINAL PLANT 1S1 '

HCLPF HCLPF MAINE-YANKEE-

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.21G

,27G CATAWBA

.15G

.3G

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.15G

.28G*

• REVIEW IS NOT COMPLETED YET, O

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INDIVIDUAL PUWT EXAMINATION FOR FIE WHY IS INDIVIDUAL PUWT EXAMlHATIQ1 FOR FIE ECESSARY?.

INDUSTRY PRA'S HAVE SHOWN FIE C0llTRIBUTES TO UP TO 50-60% OF CORE El.T FECLIENCY EXISTING CRITERIA DO NOT ADEQUATELY ADDRESS &RTAIN FIE ISSUES d

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J rfRF TW%GE FREQUENCIES FRCN PRAs EAN, CDR$ IW%GE FREGEICY/RX-YR

FIRES, STATION FOR FIRES FOR ALL IlliTIATORS FRACTION Z101 1-2 1.8 E-6 5.7 E-5 3%

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INDIAN PT 2 1.4 E-4 4.7 E-4 30%

i INDI Ai! FT.- 3 9.6 E-5 2.3 E 40%

BIG ROCK POINT 2.3 E-4 9.8 E-4 23%

LIERICK 2.3 E-5 4.4 E-5 55%

SEABROOK 2.5 E-5 2.3Ei1 12%

OCONEE-3 L.0 E-5 2.5 E-4 4%

MILLSTDNE-3 4.8 E-6 7.0 E-5 7%

PEACH BOTTOM 2.0 E-5 1.0 E-4 (LLNL) 19%

2.7 E-5 (EPRI) 72%

SURRY 1.1 E-5 1.7 E-4 (LLNL) 7%

7.7 E-5 (EPRI) 15%

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FIRE PRA o

ALTERNATE' FIRE IPEEE METHODOLOGY ic-)'"T<'.,e,c (BEING DEVELOPED BY NUMARC)

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e HIGH WIND, FLOOD AND OTHER EXTERNAL HAZARDS MAJOR SOURCES:

HIGH WIND,-TORNADO, TORNADO MISSILES, FLOODS, EXPLOSIONS,.T0XIC GASES, FIRES, AIRCRAFT AND SHIPPING ACCIDENTS 4 b iS d m,b +

g WIND:

LOSS OF 0FFSITE POWER METAL SIDING STRUCTURES STRUCTURES.WITH THIN CONCRETE WALLS AND ROOFS TALL. STACKS OPEN TANKS i

g/m-FLOOD:-

RIVER, C0ASTAL AND LAKE SITE PLANTS INTENSE LOCAL PRECIPITATION i

OTHERS:

INDUSTRIAL / MILITARY FACILITIES TRANSPORTATION (AIRCRAFT, TRUCKS, TRAINS, SHIPS, PIPELINES) 1 15 -

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POTENT 1 AL IPEEE APPROACH FOR HIGH WIND, TORNAD0ES AND l

OTHER EXTERNAL HAZARDS f;

i POTENTIAL APPROACH l

PLANT SPECIFIC HAZARD DATA AND LICENSING BASESk SIGNIFICANT CHANGES RELATED TO SITE SINCE NT0L SCREENING BASED ON VULNERABILITIES IDENTIFIED IN PAST PRA'S (E.G., TALL STACKSF INITIALSCREENINGBASEDONHAkARDFREQUENCY EVALUATION OF PLANT RESPONSE FOR SIGNIFICANT HAZARDS, MARGIN ASSESSMENT, OR BOUNDING ANALYSIS PRA-o i

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O-O SCHEDULES o

STEERING GROUP RECOMMENTATION IN WINTER, 1989 o

NEED INTERACTION WITH NUMARC 9

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