ML13322A595

Summary of 790919 Meeting W/Util in Bethesda,Md Re High Energy Line Break Inside Containment,Sep Topic III-5.A

ML13322A595
Person / Time
Site:	San Onofre
Issue date:	09/26/1979
From:	Jabbour K Office of Nuclear Reactor Regulation
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
TASK-03-05.A, TASK-3-5.A, TASK-RR NUDOCS 7910160353
Download: ML13322A595 (15)

Text

kgR REGq 0

SoU UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 Docket No. 50-2-9 MEMORANDUM FOR:

Dennis Crutchfield, Chief Systematic Evaluation Program Branch, DOR FROM:

Kahtan Jabbour Systematic Evaluation Program Branch, DOR

SUMMARY

OF MEETING WITH SOUTHERN CALIFORNIA EDISON ON HIGH ENERGY LINE BREAK INSIDE CONTAINMENT FOR SAN ONOFRE (SEP TOPIC III-5.A)

Representatives of the Nuclear Regulatory Commission (NRC) and Southern California Edison (SCE) met in Bethesda to discuss the Systematic Evaluation Program (SEP) Topic III-5.A, "Effects of Pipe Break on Structures, Systems and Components Inside Containment", on September 19, 1979. A list of attendees is provided in Attachment 1.

Following brief introductory discussions of the criteria for postulating break locations and break scenarios, SCE representatives made a presentation on the San Onofre plant layout and identified the high energy lines and safety related equipment inside containment.

Their presentation is contained in a preliminary report (Attachment 2) to be utilized primarily to illustrate the philosophy and methodology of the approach to be used by SCE in evaluating the effects of pipe breaks on structures, systems and components inside contain ment. SCE representatives also stated that they will use either the mechanistic or effect oriented approach for each separate pipe run.

To date SCE has utilized the effect oriented approach (for evaluating the high energy line break effects) described in the attachment to the NRC letter to KMC dated July 20, 1978. This approach postulates a high energy pipe break inside containment near safety equipment and analyzes the capability of the remaining systems to safely shutdown the reactor. Licensee representatives discussed the interaction of the breaks (pipe whip and jet impingement) with the safety systems required to shutdown the plant.

The NRC representatives requested the licensee to study the pipe whip and jet impingement effects of the breaks postulated under NRC TASK ACTION PLAN (TAP) A-2 and to consider these effects for postulated breaks in the primary coolant loop at locations other than those required in TAP A-2 using either the mechanistic or effect oriented approach. The NRC representatives agreed to further review the interface between A-2 and SEP Topic III-5.A to assure that there is no unnecessary duplication of effort.

The licensee stated that the consideration of a whipping pipe as having sufficient energy to pdstentially rupture an impacted pipe of equal or greater nominal pipe size and equal or greater wall thickness is an 7910160 35 3

-2 escalation of current criteria. The NRC representatives agreed to attempt to resolve this concern in the near future. However, the licensee agreed to consider the effects of jet impingement loads from a ruptured pipe on a pipe of equal or greater nominal size.

The NRC representatives made the following statements concerning the effect oriented approach:

1.

The safety objectives are:

A.

To maintain a coolable core geometry following any postulated break.

B.

To maintain the capability of safe plant shutdown (definition of safe shutdown consistent with that of safe shutdown reviews).

C.

To maintain containment integrity.

2.

It is of utmost importance that the consequences of each pipe break scenario be fully recognized and understood before a decision would be made on any proposed resolution.

The staff further stated that pipe breaks should be considered at locations close to safety related equipment (as stated in the enclosure to the NRC letter to KMC dated July 20, 1978).

Consideration must be given to the effects of larger pipe damaging smaller pipes and causing multiple failures of piping, jet impingement and single failure.

The methods of calculating the effects of jet impingement are discussed in the NRC Standard Review Plan (SRP),

Section 3.6.2 or in the proposed ANS-58.2 (ANSI-N176) dated January 1979.

Other types of enveloping solutions may be generated by the licensee and will be reviewed by the NRC staff.

The single failure criteria to be used by the NRC staff in their review will be that of ANS-51.7, Draft 4; Rev. 1, November 1975, which states that the most limiting single failure will be taken in addition to the initiating break and its effects.

The most limiting single failure can be taken either as a single active failure in the short-term or a single active or passive failure in the long-term.

Short-term, long-term and the nature of the passive failure are defined in the standard.

3.

General Design Criterion No. 17 and SRP Section 3.6.1 (BTP APCSB 3-1) will be employed relating to loss of offsite power; however, on a plant-by-plant basis, operational experience will be considered.

9 9

-3

4. Credit for operator action will be considered on a case-by-case basis once the scenarios have been developed.

The licensee requested that augmented in-service inspection (ISI) be consi dered as a means to mitigate the consequences of the postulated pipe breaks where retrofitting or adding restraints is impractical.

At the conclusion of the meeting, the licensee approach and schedule for resolving this topic were discussed. The NRC representatives expressed concern about the schedule delay in developing postulated break scenarios.

However, they agreed to discuss the revised schedule at a later date.

Kahtan Jabbour Systematic Evaluation Program Branch Division of Operating Reactors

Attachment:

As stated

DISTRIBUTION FOR MEETING SUMMARIES Docket NRC PDR Local PDR Central, Fil es-TERA SEPB Reading NRR Reading H. Denton E. Case L. Shao D. Eisenhut R. Vollmer W. Russell B. Grimes T. Carter T. Ippolito R. Reid G. Knighton V. Noonan A. Schwencer D. Ziemann D. Crutchfield G. Lainas J. Scinto, OELD OI&E (3)

ACRS (16)

Licensees NRC Participants D. Knuth KMC R. Schaffstall, KMC W. Moody, SCE H. Smith, SCE "i

n L

J. Rainsberry, SCE K. Jabbour, NRC P. DiBenedetto, NRC R. Kiessel, NRC C. Hofmayer, NRC J. Shapaker, NRC

ATTACHMENT 1 LIST OF ATTENDEES SEPTEMBER 19, 1979 Southern California Edison W. Moody H. Smith J. Rainsberry

KMC, Inc.

R. Schaffstall NRC K. Jabbour P. DiBenedetto D. Crutchfield R. Kiessel C. Hofmayer J. Shapaker

AGENDA SEP TOPIC III-5.A, HIGH ENERGY PIPE BREAK INSIDE CONTAINMENT SAN ONOFRE NUCLEAR GENERATING STATION UNIT 1 SEPTEMBER 19, 1979 I. Introduction II. General Approach III. Essential Systems IV.

High Energy Lines V. Program VI.

Conclusion

GENERAL APPROACH

1.

Identify Essential Systems

a.

Safe shutdown

b.

Emergency Core Cooling Systems

2.

Identify High Energy Lines

3. Identify Potential Impactees

4.

Evaluate Effects of Breaks

a.

Coolable Core Geometry

b.

Safe Shutdown Capability

c.

Offsite Doses Less Than 10CFR 100

EVALUATION ASSUMPTIONS

1. Lines 1" Diameter Not Analyzed

2. Piping Boundary is First Normally Closed Valve, Check Valve, Safety/Relief Valve or Valve Capable of Auto Closure

3.

Single Active Failure

4. Pipe cannot Damage Pipe of Equal or Larger Size With the Same of Greater Wall Thickness

5. Reactor Coolant System Breaks To Be Evaluated As Part of Generic Issue on Asymmetric LOCA Loads.

6. Use of All Plant Systems to Shutdown is Acceptable

7. Operator Actions Permitted

UMrY ESSENTIAL SYSTEMS LIST P&ID NO.

1. Reactor Coolant System (RCS) 568766

2. Chemical and Volume Control System (CVCS) 568767

a. RCS Letdown

b. Charging

c. Reactor Coolant Pump (RCP) seal water supply

3. Residual Heat Removal System (ERR) 568768

4. Component Coolant Water System (CCWS) 568768

a.

Cooling water for RHR

b.

Cooling water for RCP

5. Safety Injection System (SIS) and including 568769

a. Recirculation system

b. Sphere spray system

6. Main Steam System

- 568773

7. Main Feedwater System 568779 8* Instrument Air System

Page 1 Rev. 1 9/14/79 A.

(INMIlMEN1' ISDIATICN PHASE**

Buipnet FUnctian Nuter Name in Accient

1.

CV-40 Instrunant air exhaust Isolates contairnmet control valve SV-19 Solenoid valve fcr CF-40 Actuates CF-40

2.

CV-116 Sphere pressure equalizatial Isolates ontairent valve SV-127 Solenoid valve far CV-116 Actuates CV-116

3.

CV-146 Sphere vapar sanple antrol Isolate crntairment valve SV-1212-6 Solenoid valve fcr CF-146 Actuates CV-146

4.

CV-147 Shere vapar sanple (return)

Isolates ontainment cntrol valve SV-1212-7 Sblenoid valve for C -147 Actuates CV-147

5.

CF-202, 203, 204 Ioop A letdown isolaticn Isolate antainment valves Solenoid valves fcr CV-202, Actuates CV-202, CV-203 and 203 and 204 CV-204

6.

CV-102, 104, 106 Spere stup pmp discharge Isolate containment on SIS.

valve, reactr coolant drain pmp discharge valve, reactor oolant drain tank vent isolatin valve.

SV-108, 110, 112 Solenoid valves fcr CV-102, Actuate CV-102, 104, 106 104, 106

7.

(Utlesignated)

O(ntairmant isolaticn Provide indicaticn of contain valve limit switches net isolaticn valve position.

• • Pressure and radiatian sesors hich initiate cantainaent isolaticn are located outside th antaiment sphere.

.mv.

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rx-9/14/79 B.

RU=I~

pES mNber N_

in Aciet

1. FV-850 A, B & C Safety injecticn valves Orm to pernt safety injectim flow to priary loops

2.

PD-430*,

Pressurizer pressure

• Provides autonatic initiatica PD-431*,

transuitters of safety injecticn upon P-432*,

signals fran 2 at of 3 trars PL-425**

mitters irdicating lcw pres surizer pressure.

• • Provides surveillance of RC.

3.

IT-430, 431, 432, Level trananitters Provide pressurizer level 435 indicatim

4. Iop A Ractor coolant teaperature Provide indication of reactor

1. TE 402C detectors coolant taperature for surveil lance of SIS performance.

Loop B

1.

2 41.

imp C

1.

2 422C Pressrizer

1.

'E 430 A, B, C

c s v.r 9/14/79 C.

MfUJA=I-EA Epipomt Rxrcticn

_ther Mme in Accident

1.

G45 A&B Recirlaticn pzmps Oprate to povide fcr lang term <oing subsequent to safety injecticn in DA

2.

1M7-866 A,B Racirculaticn pumsp isolatica Cpen to Line up larq-tern vales recirallaticn sequ ent to safety injecticn in I.

3.

xW-356, 357, 358 Recirculaticn lire isolatia Ol to lire up 1cr-term valves reciralatial subseqtent to safety injecticn in ICCA

4.

Iop A eactor coolant terperature Provide indicaticn of reactor

1.

JE 402C

<etectrs coant tarerature fct surveil lanc of cool down.

Imop B

1.

E 412C Imop C

1.

TE 422C Pressuizer

1.

E 430 A, B, C

5.

FI'-430, 431, 432, Pressurizer pressure trans-Provide irlicaticn of pressurizer 435 mitters pressure ard level M-430, 431, 432, Pressurizer level trans 435 mitters

6.

Fr 500, 501 Flow transnitter for mtnitrs recirclaticn flow recirculaticn fran the recirc pcrs.

Page 4 Rev. 1 I

9/14/79 D.

ErUl IH ImRI'CUIcN Blinpent Fxcticn Number Nie in Accident

1.

CV 304 Crging line to tmp A CV isolates the leap A charging inalatim valve line during hot leg recircula tim

2.

CV 305 Pressurizer aux. spray CV alliows flow of recirculatia valve water to Bmp B

t leg.

3.

ICV 430C and 4308 Pressurizer sray valves lolates lCp A ard B cold legs durin hot leg recirculaticn

4.

Loop A Reactor coolant taqperature Provide indicatic of reactor

1.

'E 402C detectors colant teperature fcx surveil larce of cool down.

Ioop B

1.

2 41X imp C

1.

E 422C Press izer

1.

E 430 A, B, C

Rev.

1 9/14/79 E.

RESIMUAL EM RRI07AL BHASF"

.Nurber ame in Acident L

G-14 A&B Residual heat removal pu Cperate to provide fcr kcrq termn coaling subsequent to safety injectimi in LE

2.

?(-813, 834 Residual heat removal pres-Open to line up Irrq-term sure inter1cck valves coling subsequnt to safety injectim in fss

3.

I07-814, 833 Residial heat removal.

Open to line up rg-term isolaticn valves stbequt to safety injeticn in HE

4.

IDP-602 Residual heat exdanger egulates flow through the flow cntrol valve residual heat exdhangers daring dzring 1n3q-term coling abeqtnt to MSB

5.

F'-602

.BR f1w transmitter Provide indicaticn of M flcw

6.

Icop A Reactor coolant tetperatre Provide irticatiCn of reactor

1.

'E 402C detectors oolant terperature for surveil Jance of cool in.

Icp B

1.

E 412C ioop C

1.

'E 422C Pressurizer

1.

'E 430 A, B, C

7.

iVT 822 A&B MR heat exrcancger isolatii Provide fLow path to M heat valve exdhangers rot required pst-ILIA.

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~Page 6

9/14/79 Rpipaent Rater Mme in Acicnt

8.

E 600 REE hot ad cld leg MRnitor RR perfonaerc

. E 601 A, B tapeature transuitter

9.

RV 206 R pressure relief valve Provide overpressure protec tiai during RR ceratial F.

CNMDEN1 SPRAY L

C-82, 114 Spere spray cotrol valves Reduce fission product ornentration ard contairnent pressure durir r

ard cartairimrt pressure during KO SV-118, 128 Salerid valves for CV-82, Actuates CV-82, CV-114 G.

MIS-CEIZANEIJ

1.

Ir-450, 451, 452 Steam gearator level Mzitor steam gerator transnitters inventory

2.

ICV1115 A, B, C Reactor colant a.mp seal Provide seal and colirq water water control valves to naintain seal integrity

3.

IV 1112 and RCS letdwn control valve Col <bwn ard depressurize asaciated SV Ps post-MtB

4.

CV 530, 531, 532, FCS safety and pressure Provide overpressure protection 533, 545, 546 valves for ICS