ML20085J786

Integrated Safety Assessment Program PRA Public Safety Analysis Rept

ML20085J786
Person / Time
Site:	Millstone
Issue date:	05/25/1995
From:	Rothert J, Sunil Weerakkody NORTHEAST NUCLEAR ENERGY CO.
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NUDOCS 9506230021
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L**' INTEGRATED SAFETY ASSESSMENT PROGRAM i

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PRA PUBLIC SAFETY ANALYSIS REPORT  :

ISAP Topic No. n/a Analysis Rev. No. O Project Assignment No. n/a Rev. No. O Project Title MP2 EBFS Single Failure Resolution Project Engineer Tom A. Doyle i

Plant / Unit Millstone Unit 2 Original PRA Analyst J. K. Roth Date 05C3/95  !

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Reviewed by S.D. Weerakktxr Date 05/25/95 Updated by Date _

Reviewed by Date Change in CDF 0.00E-0/vr Risk Reduction (M R/yr) 0.003 M-R/vr Benefit (5/yr) $3.00/vr Analysis / Update Summary The scope of the MP2 EBFS Single Failure Resolution proposed project presented several options.  ;

The proposed project options must be viewed carefully with respect to their impact upon public j safety. The resolution of the identified single failures of MP2 EBFS does not preclude core damage from occurring. Also, the single failures only result in degraded system function and not total system failure. .

l Therefore, the conclusions of this analysis is that the present system has " negligible risk significance /public safety impact," and that expensive modifications to the system are not justified.

However, the following compensatory measures are recommended for AC-11 and AC-01 to maintain functionality of MP2 EBFS in light of the single failures associated with the system:

1. AC-II: Revision of procedure (s) to trip fans F-34A/B/C or manually close 2AC-08 in the case of a stuck open AC-11 post LOCA.

2. AC-01: Revision of procedure (s) to manually start F 25A (if motive power is available) post LOCA. However, we would like to note that the risk significance of this single failure is much less significant when compared to with the risk significance of 2AC-11.  ;

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i 9506230021 950606 1 i' PDR ADOCK 05000336 P PDR

r ISAP Tcpic X.XXX (MP2 EBFS Single Fcilures) i

.- Proposed Project The main objective of this proposed project options (Reference 1) is to resolve several identified single failures associated with the MP2 Enclosure Building Filtration System (EBFS). Reference 1 l

proposes several options to resolve the identified single failures as they relate to dampers 2AC-01 and

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2AC-11 which renders EBFS inoperable as per LER 94-040-01 (Reference 2).. I l

Safety Issue The proposed project options for EBFS is to resolve the safety issues associated with identified single failures with dampers 2AC-01 and 2AC-11 which renders EBFS inoperable per MNPS-2 FSAR and Technical Specification 3/4.6.5. The operability concern of EBFS stems from its inability to meet the requirements set forth in these two documents.

The identified single failures are for post Loss of Coolant Accident (LOCA) events when EBFS must t take suction on the enclosure building and draw it down to a vacuum. By drawing down the enclosure building any potential radiological releases from containment as a result of a LOCA are processed through EBFS. The identified single failures impact EBFS from performing this function '

and thus resulting in the potential for a public safety concern.

The single failure of EBFS for 2AC-01 is the loss of facility I containment isolation actuation signal (CIAS) coincidental with a LOCA. With the failure of facility 1, CIAS 2AC-01 will fail to close and fan F-25A will not automatically start. EBFS fan F-25B has insufficient capacity to draw the required vacuum with 2AC-01 open; thus, FSAR and T.S. requirements to limit off site doses could potentially be exceeded post LOCA.

The single failure of EBFS for 2AC-11 is its mechanical failure to close coincidental with a LOCA. -

2AC-11 is upstream of fans F-34A/B/C which discharge into the MP2 stack. This allows a direct potential radiological release path from the enclosure building to the atmosphere. Thus, FSAR and T.S. requirements to limit off-site doses could potentially be exceeded post LOCA.

Analysis of Risk Impact .

This analysis evaluated the public risk impact for EBFS failing to perform its design function.

The failure of EBFS does not result in an initiator which can result in core damage nor is the availability of the system credited in preventing core damage. Therefore, total failure of EBFS has no effect upon the MP2 core damage frequency (CDF).

EBFS is solely designed for post LOCA initiating events to potentially limit radiological releases from the enclosure building to the public. Therefore, the public safety cost / benefit is associated with the system being able to mitigate radiological releases following an event. ,

I To calculate the public safety cost / benefit the following assumptions were made:

The MP2 total LOCA (all LOCAs except SGTR) initiating event frequency (8.25E-3/yr, i Reference 2) was determined to be representative of the demand for EBFS. However, to get ,

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ISAP Tepic X.XXX (MP2 EBFS Singla Fellures) significant radiological releases fuel damage must occur. Therefore, the total CDF contribution of LOCAs for MP2 of 6.0E-6/yr. was assumed as the demand frequency of EBFS and wherein its failure could result in potential radiological release .

Of the total LOCA CDF, 60% of these events were assumed not to result in containment i failure (Reference 2). For these events EBFS would be required to mitigate releases to the public. For the remaining 40% containment failure beyond design basis would occur and that  ;

EBFS would be ineffective to mitigate radiological releases. 1 Coincidental failure of either CIAS or 2AC-Il was conservatively assumed to be 0.01.

Assuming that other containment heat removal functions are available (CAR fans and/or containment spray) the MP2 radiological consequences are 7.5E+4 person rem per core damage (Reference 3).

Thus, the calculated public safety benefit from implementation of any of the proposed project options is calculate as;

$3.00/yr = ((6.0E-6/yr)(0.6)(0.01)(7.5E+4preson-rem)($1000/ person-rem))

The cost / benefit of htis issue resolution was assessed by an independent method. MP3 had a similar issue with regard to SLCRS (Reference 4). Using infonnation from that analysis and extrapolating the data for MP2 the benefit was calculated as $0.40/yr (see attached representative figure 1).

Where the equivalent radiological release for MP2 as a ratio of MP3 data for containment fails case is approximately 220 person-rem year.

~220 person-rem /yr = ((3.4E-5/yr)/(5.5E-5/yr))((40)/(20))((2700)/(3425))(220 person-rem />T) where; 3.4E-5/yr is the MP2 CDF 5.5E-5/yr is the MP3 CDF 40 is percentage of MP2 containment failure 20 is percentage of MP3 containment failure 2700 is MP2 thermal MW rating 3425 is MP3 thermal MW rating 220 person-rem /yr is MP3 calculated radiological consequence due to containment failure as a result of core damage Therefore, the equivalent design leakage associated with the MP2 is 0.04personerem/yr.

Result The proposed project options to address MP2 EBFS single failure concerns have been determined by  ;

PRA to have negligible public safety benefit.

Therefore, the recommended conclusion of this analysis is that the present system has " negligible risk l significance /public safety impact" as a basis not to perform potential expensive modifications to the system.

3

e j ISAP Topic LXXX (MP2 EBFS Single Failures)

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• - However; to minimize any possible consequence / risk in light of these single failures, the following compensatory measures are recommended for AC-11 and AC-01 to maintain functionality of MP2 EBFS:

1. 2AC-11: Revision of procedure (s) to trip fans F-34A/B/C or- manually close 2AC-08 in the case of a stuck open AC-11 post LOCA.

2. 2AC-01: Revision of procedure (s) to manually start of fan F-25A (if possible)in the case of a stuck open 2AC-01 post LOCA.

References

1) Memo from Tom A Doyle, " Meeting Minutes, April 6,1995, on AC-01 and 11 design problems for Secondary Containment Integrity" with attachments, DE2-95-254, April 28, 1995.

2) " Millstone Unit 3, Individual Plan Examination for Severe Accident Vulnerabilities,"  :

NUSCO 171, August 1990.

3) Memo from J.K. Rothert, "IS AP - Public Safety Attribute Evaluations for CY, MP1, MP2, '

and MP3," Rev. O, Ne-92-SAB-238, July 31,1992. .

4) Presentaion Material at MP3 Enforcement Conference on SLCRS/ABFS Operability, ,

January 11,1994.

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ALL CORE DAMAGE SCENARIOS

=

3.40E-5 =

1.4E-5 CONTAINMENT FAILS 40%

2.0E-5 DESIGN '

LEAKAGE 60 % y >

.04 PERSON REMS PER YEAR

~ 220 PERSON REMS PER YEAR COINCIDENTAL SINGLE FAILURE

(.01)

.0004 PERSON REMS 99 %

PER YEAR 1%

DOSES FROM NOBLE GASES

- $1000/ PERSON REM

$0.40/ YEAR l (POTENTIALLY BENEFIT OF EBFS OPERATION) l ]

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RISK SIGNIFICANCE OF EBFS MODS FOR MP2 Figure 1 .

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SECONDARY To uMIT 2 CONTAINMENT STACK - AC-8 ENCLOSILRE j (

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a s AC-6 AC-4 EBFS FANS START ON EBFAS ENCLOSURE BUILDING CONTAINMENT VENTILATION

, 2AC-01 AND 2AC-11 PROBLEMS PAGE I 0F 2

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ox 270 HART FORD. CONNECTICUT 06101 (203) 666 6911 L c A ' ,:'D ".".,::;.'::ll:::.1 March 1,1979 I Docket No. 50-336 l

Director of Nuclear Reactor Regulation Attn Mr. R. Reid, Chief Operating Reactors Branch #4 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

References:

(1) D. C. Switzer letter to C. Lear dated September 22, 1977.

(2) G. Lear letter to D. C. Switzer dated October 12, 1977.

(3) D. C. Switzer letter to R. Reid dated March 13, 1978.

Centlemen: -

Millstone Nuclear Power Station, Unit No. 2 Enclosure Building Design

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, In Reference (1), Northeast Nuclear Energy Company (NNECO) advised the NRC Staff

\") that the ten-inch suction line to Fan F-55 was not designed and supported seis-mically. This situation was reported as a 24-hour Reportable Occurrence, on the basis of a postulated coincident safe shutdown earthquake (SSE) and design basis accident (DBA) . Under such a circumstance, failure of this line could prevent the enclosure building filtration system (EBFS) from maintaining a 0.25 inch negative pressure in the EBF region. In Reference (2), the NRC Staff concurred with the assessment and the proposed corrective action, that of qualification of the line as seismic Category I prior to Cycle 2 operation.

In Reference (3), NNECO indicated that two additional non-seismically supported lines in the EBF region were discovered. All modifications were completed prior to the start of Cycle 2 operation, as reported.

The purpose of this letter is to advise the Staff of recent investigations in this arca, and report NNECO's conclusions.

NNECO has determined, based on further review, that the existe.nce of non-seis-mically designed and supported lines which penetrate the enclosure building is acceptable. This situation is, in fact, the original design basis as reported in the FSAR and approved by the Staff. The response to FSAR Questions 6.9 and 6.16.4, provided in Amendments 39 and 16, respectively, discusses various non-seismic Category 1 penetrations through the EBF region. This configuration was, and continues to be, an acceptable design basis. As stated previously, the basis for the Reportable Occurrence of Reference (1) was a postulated coinc.ident SSE and DBA, or DM followed by an SSE. Although such a postulate was clearly a safe method to evaluate the adequacy of the design of Hillstone Unit No. 2, it was also excessively conservative. Having recently identified certain addi-tional non-seismic penetrations, NNECO has further reviewed and evaluated the original design basis as well as current regulatory guidance in this area, and concluded that the current configuratic,n is acceptable as presented below.

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• C A thorough review of the Millstone Unit No. 2 FSAR Sections 5.3.3.1.3, i

j 5.3.3.1.4, 5.3.3.2, and 5.3.3.2.1' reveals that although the enclosure building j is designed to mitigate the consequences of a DBA and to amintain its structural  !

integrity during and af ter a seismic event, simultaneous occurrence of the two l cvents was not the design basis for the enclosure' building. I The Staf f's Safety Evaluation Report, dated May 10, 1974, states in Section 3.9  !

that "an appropriate combination of loads likely to occur" was considered in j the design of the enclosure building. The NRC Staf f did not require considera-  !

tion of a simultaneous loss-of-coolant accident and seismic event for the  !

enclosure building when' the' design was approved.  !

l As stated in General Design Criterion 2 of 10CFR50, Appendix A, only " appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena" need be considered. For example, it is clear that tornado events were not required to be combined with either loss-of-coolant events or seismic events due to the low likelihood of simultaneous occurrence. j Appendix 5.B of the FSAR states that seismic events and loss-of-coolant inci- 4 dent were considered together, but only for the purposes of assuring conservatism in specifying appropriate load combinations used in design equations of the con- i tainment structure. These two events were not combined for the purposes of conducting a consequence evaluation regarding performance of the EBFS.  !

This logical approach of not combining remote probability events to evaluate  ;

consequences is evident in a number of regulatory documents. For example, Regulatory Guide 1.117 " Tornado Design Classification" states that "it is not necessary to maintain the functional capability of all Seismic Category I structures, systems, and components because the probability of the joint i

occurrence of a low-probability event (loss-of-coolant accident with DBT or

• smaller tornado, or earthquake with DBT or small tornado) is sufficiently small".

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NNECO fully realizes that SSE and LOCA loads have long been combined by l means of factored loads for the containment structure and reactor coolant j piping system analysis. As indicated by the NRC Director of the Division of l Systems Safety, Dr. R. J. Mattoon, at the June 2,1978, ACRS meeting, the  !

reason for this conservative load combination is not clear to the NRC Staf f j at the present time. The NRC is presently developing a methodology for es-i tablishing the appropriate margins in the containment structure.and reactor '

coolant piping system design to replace the arbitrary combination of loads i from two remote probability, independent events (NUREG-0484, 'Hethodology for .

Combining Dynamic Responses").  !

h to summarize, the enclosure building structure is designed to retain structural i integrity subsequent to a seismic event. However, the EBFS is not designed to be functional subsequent to an SSE.

i The hypothetical situation of coincident LOCA and SSE is beyond the scope of the original Millstone Unit No. 2 licensing basis. Further, such a postulate

'( has insufficient technical basis, and, therefore, cannot be considered credible.

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t NNECO is not docketing this letter' to request Staff review, but rather to-clarify the misleading and excessively conservative interpretations we pre- '

sented to clarify our position.

in References (1) and (3).- We trust the above information is sufficient 1

Very truly yours, NORTHEAST NUCLEAR ENERGY. COMPANY  !

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W. G. Counsil '

Vice President ,

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RECEIVED 19 J. J. Kelley g g 3 gg}

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,ess R. M. Eacich guesegg Millstone Unit No. 2 Enclosure Building Design j

i Dy letter dated March 1,1979, setECO docketed our position regnading coincident IDCR and sst events as applied to the design of the anclosure Building. Since the docketing of that letter numerous discussions.have been held with the staff regarding the interpretation of our posittaa, the most recent Inte position is summarimod in the attacited memorandum frcan R. M. Novak to J. P. Enight, dated May 21, 1981. Por the first '

time, the staff is recommanding that the Millstone Unit No. 2 positice be accepted. The 15-day review interval identified in the attached memorandtsa has espired without comument. On that basis, the sec staff has closed out the tac (Technical Assignment control) s, e sociated with this effort. We have been unsuccessful in having the lec docket this position to us in a formal manner. The purpose of this memo is to:

o Document the current design basis for the anclosure Building as it may be faterpreted for future backfits.

o Indicate icic's basis, in part, for accepting this position. NUlt2G/CR-1889 is raferenced to substantiata the technical adequacy of this position.

Please advise if you require any additional clarification of this asttar.

308/ pag ^

cc 2. J. Mrocska C. 2. Cornelius J. P. Opeka M. P. Cass

3. A. DeBarba M. P. Sain C. J. Gla&iing P. A. Blasioli R. T. laudenat.J. R. Rimmelwright R. P. Necci

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$ NAY 22 lut MEMORANDUM FOR: J. P. Knight, Assistant Director for

,' Components and Structures Engineering FROM: T. M. Novak, Assistant Director for Operating Reactors

SUBJECT:

ACCEPTA81LITY 0F THE MILLSTONE, UNIT NO. 2 ENCLOSURE SUILDING AND ITS FILTRATION $YSTEM Late in the construction of Millstone Power Station Unit No. 2 (M-2), the staff stipulated that a further reduction in the off-site dose following a LOCA was necessary. Northeast Nuclear Energy Company (NNECO) responded ,;

by proposing an enclosure building (ES), a limited leakage steel framed structure, completely surrounding the containment above grade. This solution was found acceptable as stated in our May 10,1974 Safety Evaluation Report.

i In the later part of 1977. NNECO discovered and notified the staff that someenclosurebuildingfiltrationsystem(E8FS)lineswerenotdesigned and supported seismically. These problems were corrected prior to the start of Cycle 2 operation. However, in a follow-up letter of March 1,1979 (Enclosure 1), NNECO provided their conclusions based on additional investi-gation

" into the seismic design of the E8 and the E8FS. They summarized that,

...the enclosure building structure is designed to retain structured in-

' tegrity subsequent to a seismic event. . However, the EBFS is not designed to be functional subsequent to an SSE". This statement means, according to t

conversations with NNECO, that the IB suppott frame and the entire E8FS will be intact following SSE, but the sheet steel may tear at some locations. Thus, maintaining a 0.25 inch negative pressure in the EB (the system design objective) is not assured after the SSE.

This problem was the subject of a meeting with you held in sty office in late September 1980. At this meeting, we got the impression that the staff requirement, that seismic loads be combined with LOCA loads for a building performing a backup to containment function (such as the E8 at M111 stone-2),

may be changing. This possible relaxation in the above stated requirement by your staff was also noted at the April 15, 1981 meeting on Asymetric LOCA Loads, Unresolved Safety Issue A 2. We understand that justification for such relaxation stems, at least partially, from research reports such as j NUREG/CR-1889 which documented the calculated probability of a combined l LOCA and SSE at 1.8 X 10-12 (Enclosure 2 contains selected pages from this report).

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Based on the above, it does not seem reasonable for the MAC to require l

• NNECO to modify the EB to meet the combined LOCA plus $$E load combination i

- criteria. If you agree with the preceding statement. TAC No.11522 will be closed-out as completed and Mr. N. Romney of your staff should not charge any further review time to this TAC.

If, however, you disagree with our close-out of this issue on the M111 stone-2 docket, please provide us the technical basis and schedule for completing your appreciated. review of this issue. Yourresponsewithin the next 15 days would be l

O ~)erM Thomas M. Novak, Assistant Ofrector for Operating Reactors Olvision of Licensing

Enclosures:

As stated

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krch 1. 1979 Docket No. 50 334 l Director of melaar Reactor Regulaties Atta: att.1. Raid. Chief operating Reactors Branch f4 U. 3. Ihnelear Regulatory Cousiastes Hoshington, D. C. 20333 Referencess (1) D. C. Svitser letter to C. Lear dated September 22, 1977.

. (2) C. Lear letter to D. C. Switser dated October 12. 1977, (3) D. C. Switser letter to R. Raid dated March 13, 1978.

I Castlemen Millstone 3haclear Power station. Unit me. 2 Eaclosure Buildina Desian I la Raferonea (1). Northeaet Raclear Eaergy Company (MrECO) advised the BC 8taff .

that the tea-inch suctica lima to Fan F-35 was not destgaad and supported asie- )

alcally. This situaties was reported as a 24-hour Reportable Oscarremes, en i the basis of a postulated cetacident safe shutdeva earthquaka (SSE) and destga basis ac:1 dent (Da&). Veder each a circumstance, failare of this line ceuld prevant the enclosure building filtratica system (IIFS) from maintaintag a 0.13 iach negative pressura in the EIT region. la Reference (2), the BC staff concurred with the assesement and the proposed corrective acties, that of qualification of the line as seismic Category I prior to Cycle 2 eteraties.

In Rafarance (3) NMECO indicated that two additional apa-seianically supported lines in the EIF region were discovered. All modificatimas were completed Frier to the start of Cycle 2 operaties, as reported.

The purpose of this letter is to advise the Staff of recent investigations ta this area, and report IDfECO's conclustoas.

IBfECO has determined, based es further reriev, that the esistence of mes-eate-mically designed and supported 11 ass which penetrate the enclosure building is acceptable. This situatica 4c. Sa fact, the original design basis se reported La the FSAR and approved by the Staff. The respease to FSAR Questions 6.9 and 6.16.4. provided La a=w==ts 39 and 16 raepectively, discusses various sea-seismic Category 1 penetrations through the E5F regiaa. This seafiguraties uns, and continues to be, an acceptable design basis. As stated previsualy, the basis for the toportable Occurrence of Eaference (1) sus a postulated ceiacident 85E and Da&, or DBA followed by an 88E. Although each a postulate uma clearly a safe method to evaluate the adequacy of the destas of it111 stone Unit Es. 2 -

it tes also excessively conservative. Eavtag recently identified certain addi-tional non-saismic penetrations, 30tEC0 has further reviewed and armluated the original design basis as well. as current regulatory guidance is this ares, and A concluded that the currect configuration is acceptable as presented below. (

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A Wrough review of the Millstone Unit bio. 2 FSAE sactions 3.3.3.1.3, 3.3.3.1.4. 5.3.3.2, and 3.3.2.2.1 reveals We although the enclosure building is duissed to mitigate the consequences of a DEA and to asistain its structural

. integrity during and af ter a seismic evett, sh1 *25tts t occurrence of the two events was not the destra basis for the w1 Fours buildiat, The 8t:Jf's safety realuatica toport, dated May 10, 1974, stataa la secties 3.9

- that "as appreoriste combinaties of leads likely te occur" ans considered La

. the design of the enclosure buildias. N 3RC Staff did not require coasWra I ties of a building enclosure simultaneous when the lose-of-coolant accident and esianic event forj the desiga aus approved.

As stated in Cenatal Desien Criterien k of 10CF150, Appendia A, *aly "speropriate combinations of of the affects the natural phemosena" of eormal amed be considered. and occident acaditions with the effects For saaeple, it is clear he l I torando events were not requ! ed to be ceabined with either lese-of-coelaat l events or seismic events due to the lov 11ks11 head of simultaneens occurrence. l Appendia 3.3 of t.he FIAE states that asiaala events and lese-efw.eelsat inci-dant were considered tagethat, but only for the perpenes of assaring sesservatism is specifying tatament appropriate load combinations used sa desiga equations of the_ con-strucevrs. These two events were not combined for the purposes of

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chaducting a consequence evaluatsoa regarding perforesacs of the EIF3. ,

This logical approach of not combinias remote probability events to evaluate consequences is evident la a number of regulatory documents. For esseple, Esgulatory Guide 1.117. "Torando Desiga classification" states that "it is not necessary to maintain the functional capability of all Seismia Category 1 structures, systems, and components because the probability of the jotat I occurrence of a low-probability event (lose-of-coolaat accident with DDT or smallar tornade, or earthquaka with DRT or small tornado) is sufficiently satil".

NNECO fully realises that ESE and LOCA loads have long been combined by means of factored loads for the containment structure and reactor coolant piping system smalysis. As indicated by the IEC Directar of the Diviaism of systems ram - safety, Dr. R. J. m~n, at the June 2,1978, Acts meettag, the N "'- -Ennerrative load combination is not clear to the IRC Staff

.at the present time._ The Mtc is presently developing a au=ccAegy .n w- -

cat.11shing the appropriate margins in the containment structure and reactor coolant piping systes design to replace the arbitrary combination of leads from two remote Combining Dynamicprobability, Easponses").independent events (NU12C-0464, "Wethodology for To summarise, integrity the enclosure subsequent to a seimicbuilding event.structure is designed to retain structural be functional subsequent to an 35I. towever, the Ears is act designed to The hypothetical situation of coincident IDCA sad $51 is beyond the scope oi the original Millstone Unit so.1 licensing basis. Further, such a postulate has insufficient technical basis, and, therafore, cannot be considered credible.

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3 1ersco is est dechattag this letter to request Staff review, but rather to alarify the alaleadias and eneassively seeservative laterpretattees we pre =

seated .ia Refersaces (1) and (3), We trust the above taformattee to sufficient

- te clarify our yesittee.

Very truly peers, i

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. ,Proaability Assessment-Large LOCA-E Load Combination Program i

Project 1 Summary Report Y. sesR. c. K. Chev '

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uvennere, cA Meso Prepared for DMalon of Reactor Safety Research  ;

Office of Nuclear Regulatory Research  !

U.S. Nuclear Regulatory Commitalon Washington, D.C. 20506 NRC FIN A0133

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AnafucT This report summarises tsock performed for the U.S. Isoslear Segetatory

, ceasesien (tsc) by the tend combination progree at the 8.awrence Livermore t -

Irational Laboratory to establish a technical beats for the tsc to ese la

• reassessing its requirement that earthquake and 1arte lose-of-coolant aseident (EcCA) 1eeds be combired in the design of mustur power plants. A systematie probabilistie approeah is used to treet the random nature of earthquake and t.ransient Leoding to estimate the probability of large tcCAa that are directly and indirectly induced by earthpakes. A large IACA is defined la this report j

as a double-ended guillotine break of the primary reactor ecolant loop piping (the het lege sold leg, and crossover) of a pressurised water reaeter (pwa).

Unit 1 et the sien puolear Power plant, a four-loop Pus >1, is seed fee this i study.

l To estimate the probability of a lette tacA directly indueed by earthquakes, only f atigue erack growth resulting from the sembined offsets of thermal, pressure, solante, and other cyclie loads is sensidered. Fatigue erset growth is simulated with a deterministie fraetere seehantes sedel that ,

incorporates stochaette inpets of initial orack sise distributies, material properties, stress histories, and leak detection probability. Seesita et the simulation indleats that the probabili,ty of a double-ended guillotine break, either with or without an earthquake, is very ana11 (on the order of it*1I) . The probebillty of a leak wea found to be several erders of magnituds greater than that of a ooppiste pipe reture. A limited i

investigattan involving engineering judgment of a double-ended guillotine I

• break indirectly induced by an earthquake is also reported.

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l ghe Code of rederal Segulations requires that structures, systems, and g .

eosponents important to the safety of nuclear power plants in the Dnited sutes he designed to withstand appropriate sonbinations of effects of natural l phenomena and the effects of normal and moeident oseditiona.1 31stortoally, e

the U.S. Nuclear Regulatory Consissian (IsIC)==through Segulatory Guides, g

regulations, branch technical positions, and the Standard Review Flee--has e

required that the responses te various moeident toeds and toeds oessed by natural phenomena be oombined in the analysis of safety-related structures, eyatems, and components.

• Designing safety related structures, erstaas, and sosponents to withstand the combined effects of an earthquake and a large loso-of-ooolant aseident

. (14CA) is one such land combination requirement that has been implemented ley

, the neslear ineastry for more than ten years in the design of commercial l auelear power plants. The combination of the most severs IACA lead with safe shutdom earthquake (ssr) loads was set controversial entil about five years ago when the postulated tecA and ass leeds were both laeressed by a feeter of two or more to account for such phonemena as asynnetrie blowdown la suas and I techniques to better define the leading were developed.

As a result of this change, the sembination requirement beoane more  !

difficult so taplement, particularly in the design of remotor pressure vessel  !

internal and support systems. por future plants, the change brought with it l the prospect of lacrossed construction soets. Additionally, the load oombination requirement raises the issue of whether design for sattene loeds

, will result in reduced reliability during normal plant operation. For  !

l emample, present solante design methods tend to reesit la stiff systems and I more ogports when additional strength is provided for the earthquake loading. Because a stiff eystem is subjected to greater eyelle thermal stress than a flexible one under normal thermal operating toeds, reliab!!ity is reduced under normal conditions.

Faced with these designe opet, and safety issues, the nueleer industry petitioned the IstC to reconsider its design requirement. Free a safety ,

viewpoint, ooets should not be a f aster in changing design requirements. The e

soets of seeting design requirements are industry's responetbility. Bowever, 1

1 .

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s for estating planta to meet the revloed toeding definition and aise satisfy the sonbination requirement, modifiestien is almost unavoidable. Certain i

,, plants een be feasibly modified, but other plants are not feasible to modify, l C

and they present a diffloult probles to the unc. N solution saa be either to altos sentinued operation without modifications, aballenge the safety of

, sentinued operation without modifloations, or solisit a technical basis for l

. reassesslag the design requirement.

l Industry's approach to the probles has been to esente safety by justifying that the sonbination of events is unlikely. per the'aut to essept industry's justifloation and change the destyn requirement, independent l

oentirmatory research is necessary. One such study has already been made for the Whc by Battelle Columbus Laboratories.2 N Sette11e researobers used a =

deterministle approeoh to assess the likelihood ed a break ooourting la a sold '

iogpipeofapWRplant. The uork reported herela assesses the het leg, esid leg, and erossover of a FWR plant. The appreech used is probabilistie. Both efforts will be considered by the Mac in assessing the nuclear power industry's submittal and request.

OBJECTIVE The objective of this study is to estimate the probability that a large

!acA and an earthquake occur simultaneously. This Laformation will be used by the Inc to reassess the requirement that earthquake and tacA loads be combined. If a IACA and an earthquake are independent events, the probability of simultanews oceurrence is espected to be very low. sowever, if an earthquake can cause a IACA, the probability of simultaneous occurrence may be signifloant. Thus, this assessment considers only !ACAs that are directly and indirectly induced by earthquakes in addition to normal and abnormal plant operating loads.

i scop 1

In phase I, wo limit our investigation to deteralaing the probability od a large IAcA induced by earthquakes for a pressurised water reactor (ptet) plant. A large 14cA is defined as a double-ended guillotine break of the primary reactor coolant loop piping--the hot leg, cold leg, and crossover.

Such pipes typtoally have outside diametere of 30 inchee or more, and have 2

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I wells that are appreatmately 3.5 laches thisk. This evaluation is lialted to the rupture of such large piper because they will generate the moet severe lacA loade, which, when enahined with 883 loads, present the design and l retrofit preM ans disenssed abees. -

i ,

we recognise that the break of a analler pipe say be more probaMe, and that such a mall IAcA asy pose larger risks to the plaat. sowever,'for phase

' 1 wo limit our esope to the large tacA defined above in order to address the immediato unc need for eenfirmatory reseerek. We believe that the models and ooeputational procedures developed for the large tcCA man be entended to the

! asees .ent of s.au n tocha doing soboe , vent ,h..es o ou, et.dr.

Only f atigue ersek growth resulting from the sembined offsets of thermal, pressure, seisste, and other eyelle loads was considered as the neohanism leading to soeplete pipe rupture as a direet sensequence of earthquakes. The water hasuser seehanies was not considered besasse it has never been observed in pun primary systems. Likewise, stress sorroeien is another plausiMe mechantes, but it was enoluded from consideration because the coolant water ehesistry in pwks is such that stress corroeien problems have mot been ebeerved.

A lialted investigation involving engineering judgment ed a double-ended guluotine broek indireetly induesd by an earthquake is also reported.

Barthquake-indweed indirect eenses such as fauing eraaes, mechantoal, electrical and streetural f ailure, as won as fire, emplosion, and missiles f are considered. The emphasis of this work was to identify sourses and establish the ground work for a more thorough evaluation in a e:M: pat g phase.

Therefore, the preliminary resulta presented in this report are

, limited.

l

• tmit 1 of the fien Noelear power plant itten Unit 1) was selected as the t demonstration plant for this study. The results and conclusion are app 11eable only to tien Unit 1 at this times that is, no attempt has been made to eatend our findings to other plants during phase !. Bowever, the methodology developed for the evaluatism is an advanced computational tool. It aan be applied in future evaluations to the break of reactor coolant pipes, large oc anau, to other pwn and boiling water rosetor inset) plants, or to general piping reliability assessments.

such studies are clearly beyond the scope of ,

the work reported here, but they may be part of future phases of the program. -

1 e

ArracACE The surrent practice of ocasidering these dynaale ownts acting ,

concurrently has generally been based on conservative engineering judgment that has not addressed the feet that the postulated 1cCA and earthquake leads are randce events. Amplitude, duration, f requeney sentent, time of escuronoe, and time-phase relationship are random and stochaette la nature. Thus, a systematic probabilistie assessment is necessary before a technical basis for .'

an appropriate combination requirement can be developed.

A multiphase, systematie probabilistie approach was seed to treet the random nature of earthquakes and tacAs. In reaching our Phase I objective, we took the following steps .

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l e considered many nochantaas that een lead to a pipe failure as a direet .

result er an earthquake, but concluded that fatigue creek growth resulting f rom the combined eff ects of thermal, pressure, seismie, and .

I ether cyclie loads has the highest potential to lead to sosplete pipe rupture. In particular, t.he water basseer effect was not considered ,f l

because it has never been observed la pas primary syntaas. .

e Modeled f atigue stack growth with a deterministie fracture mechantos model that incorporated stochastie inputs of initial crack sise distribution, setorial properties, stress histories, and leak detection probability.

e Denloped structural models to caloulate seismic stresses and nonsolante stresses induced in the piping by dead weight, pressure, thermal expansion, and transients. Steady state vibrational stresses and residual stresses were also considered but found to be insignifloant for growing f atigue oracks.

e calculated the probabilities of pipe leaks and breaks during the plant's lif e by inputing to the fracture nochanics model the results ef the stress analysis and estimates of the ereck aise distribution, material properties, and crack and leak detection probabilities.

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I e Estimated the probability of a directly indosed large SACA beoed en the site-specifie seismis hasard.

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e performed 11 sited sensitivity studies en the inputs to the stress

. analyses, f reeture seehantes model, and estination procedure.

e feed engineering judgment to estimate the probability of a large toch ladueed indirectly by earthquakes. streeteral, soebentest, and I eteettiest failures, as well as esplosions, fires, and missile insidents caused by earthquakes were considered. The emphasis of this work was ts identify neuroes and establish the ground work for a more I

thorough evaluaties in a subsequent phase. Therefore, the pre 11almary resulta presented in this report are limited.

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Assumpticas att neesssary to staplify the comptes assessment. ghese assumptions and simpitfisattens refleet engineering judgment that is bened en

, information available free the literature and from preliminary analyses. ghe asseptions for the d,tractly induced IACA prehten are listed belev, with the most,beste ass o ption listed first.

e Failure results from f atigue ersek growth of eraok-like defects that are confined to the girth-welded butt joints. Note that these tracks are modeled in the alreumferential orientation beesuse of its referones to the postulated doehle-ended pipe break, Emngitudinal eracks, theogh taportant to the leak assessment, will not reesit in seek a guillotine break and are, therefore, not eensidered.

e crack sise distribution enn be adequately sharacterised, despite the lack of data en large stacks. The ta!! et the stack slas distrihetion (i.e., the large oracks that are present from the enset of erack growth) plays a major role in determining the prehability et a 14CA.

trafortunately, there are no esta en large eraekst thes, ereek distribution data generated from analler erask sises must be extrapolated into this regime. Several distributions were considered,3, el and the most conservative distribution (that of

, Marshall, Ref. 3) was used, e Based on ebeerved pipe cracks, the erack shape is semiellipticals the shape is malatsined during erset growth; and the length-to-depth aspoet ratio is two or sure. Only surface oracks are evaluated. aseed en equivalent crack surf ace area and erack length, a surf ace sente111ptical flaw has a stress intensity 1.4 to 2 5 times that of the embedded flev. When this is east in terms of fatigue behnior, we find that the surf ace flaw will grow at a rate of 4 to 50 times that of the subourface flaw.

e no distinction is made between the shop or field welde for either material properties er ereck sise distribution. Data on fatigue and tensile properties of welded and unwelded 318 stainless steel show little verlation in mechanical behavior.6,7,8,9 ,

6

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e the initial erack sine distribution is independent of locaties within a joint and from joint to joint.

e At most, one orack is sedeled in each joint, and crack lateraetten is not eensidered.

8tuttiple eraeks are ignored because the probab(18ty of having esactly one essek in a joint is appresiastely 0.p9, whereas the

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probability of having two er more eraeks in the mane joint is less thes 5 a 14"3 These apprealaattans are heaed on the probability of having a creek in a unit votame equal to 10*8/in.3 (an adaptation '

free Refs.10 and 11) and a weld volume of appresinately 143 la.3 per weld joint.

• e suber!tical crack growth results free fatigue. St.ress eerrestem i erecking is not eensidered booemee it has not been observed in Pam primary coolsat piping.

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e A Paris model esseribes steek growth.6.12,13,14 Crack retardation stemming from pulsea ed high-strees eyeles is not eensidered, nor is the enhanced fatigue behavier resulting from large values et etress intensity. The effects et material variation, weld properties, and environment the Paris equation. are sooounted for in the distribution of the seeffielent to trhan an inside entface orack grow through the i

• pipe well, its length on the outside surface is conservatively set equal to the.inside length. Therefore, the minimum through-wall erask to at leest twice the pipe thiekness.

e only cracks in the of reumferential orientation are eensidered.

Probable crack toestions that are not considered in this study include longitudinal welds in elbows and pipe sections, nos,ste oorsers, binetallie transitten joints it.o., the rosetor pe,esauro vessel to safe-end welds), and base metal. Enosyt fee the hinetallis transitten joint, these creek locations will not result la the double-ended pipe break.

Data for f atigue of the bimetallie jeant are uneve11ahle.

sowever, if the 314 etainless steel fatigue relation is a reemenable approximation of the binetallie property, then the systen fallare

• probabilities should not change significantly.

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e stresses used in this evaluation include pressure, thermal esponsion, deed weight, operating and ahnstmal transients, and seismis events.

. Conservative values of desip temperature and pressure shanges and

'r frequeney are esed la palestating stresses. Vibration stress segnitudes are very low, and for ereck sises et interest seemit la a

, stress intensity below the thresheid value. Besidual suesses are

.

• found to be compressive en er very near the inside surface, and the i suess intensities, evaluated free them are nogetive. mesidual suesses are not ineladed'because (1) the negative stress intensity would retard eteek growth, and (3) there is a eartain amount of anoortainty assostated with residual suesses and how they redistrhte as a function of ereek growsk. ,

l e Desi p , construstion, and assembly errors are not estimated. ghe primary piping is subjected to high quality assurance testing and if grees design er construetten arters exist, they would be found"during systes eheckout and hydrotesting.

e support stiffnseses and structural damping values are estimated based on the best avellable data. All esmponent supports and enabbers are assumed ts be in good working order. The soil model and seismis hasard surve are both based en site-specifie estimates.15,16 .

i e Transients oesur as a stationary poiseen process reflecting the sten Unit 1 operating history for its seven years of operation. Entthquakes characterised by a peak horisontal ground acceleratten less than 0.e7 g l are assumed to affect steek growth negligibly. . ghe law stresses '

ealeulated bear this set. The naminum free-field peak soosleration ,

that aan be generated at the sien site from the available data base is I

'0.85 g (5 a sar). .

1 e both leak and laste LCCA fa!1eres are eensidered. Cracks which grow through the piping us11 lead to leaks of greater than 3 gym based on the sinimum ersek length of twice the wall thickness. These leaks are assumed to be detected and repaired. The orack sise distribution is set squal to the initial sise distrNtint af ter sold repair.

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• e A large WCA can occur if the lead-oontrolled stresses are high enough 1 to sever the pipe. It is assumed that dead weight, pressure, and 4 seismie stresses are lead sentrolled. Soake' that soeur and lead to a.

1ACA during a seisade event are not eyeleated--enly the flaal result (1eek er ECCA) af ter the earthquake is considered.

o Pipe severance results free met section plastle instability, which I

escurs when the not secties strees essoeds the average flew stressa agg,,= (#7 + p,)/2. The flew stretS is 8ssum04 te he mereally

} distributed. Approximation of the J-integral and the applied tearing 8

y!sid oonsidershly larger ereck sises.

a e the simuistion is ended when the first operstist basta er larger earthquake securs. '

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3 SID91ARY OF RESULTS s

mesults of the simulation indicate that tk probability of a large icCA,

' either with or without an earthquake, is very anali. The probability of a large tcCA induced by plant transients given the condition that no earthquake occurs charing the 40-year, plant life is estimated to be 1.6 a 10-12, solamic stresses tend to increase this probability; that is, the probability of a large 1DCA induced by plant ter.nsients and earthquakes is estimated to be 1.8 a 1012, assuming the tk plant is ht h af tar a prating basis or larger earthquake. The probability of an earthquake and an earthquake-induced large EDCA occurring staultaneously sharing the plant's life is about 4 s 10*13 These findings are consistent with othr studies of

• this nature.

• These teruits are best estimates based on the approech taken and the assumptions made. A limited sensitivity study on input variables was perfoceed. Input verf ables were assigned incredible values to determine how the results vary for the following cases e No preservice laspection e No preservloe bydrostatic proof test e No leak detection e stodifiod Marshall initial crack distribution e variation of aspect ratio of the length and the depth of the initial crack e Increase the 5 a art seisale stress by a factor of 3.

• The limited sensitivity study indicatas that leak detection capability, preservice inspection, and proof testing tend to reduce the probability of a ICCA by less than an order of magnitude. Increasing the percentage of cracks with large aspect ratio increases the large ICCA probability sigolficantly; however, it has little influence on the leak probability. -

Results of the simulation indicate that an earthquake has almost no effect on the leak probability, which is estimated to be 8 a 10" during the plant's life.

In view of the extremely low probabilities reported above, it oma be concluded that a reactor coolant loop pipe break as a result of fatigue orack growth is highly unlikely. The probability that an earthquake and a ECCA induced by an earthquake occur simultaneously is less. These conclusions toply only that a reactor coolant loop pipe is unlikely to break as a direct j 10 -

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• e result of the combined loads from normal plaat operation and an earthquake.

Other esternal earthquake-induced sourose--such as fires, explosions,

, alssiles, and structural, mechanical, or electrical failures--can potentially cause a pipe break as an indirect result et an earthquake. A lialted investigation of such 14CAs indirectly laeaced by an earthquake revealed that most sources do not result la a IACA which falls withia car definition et a

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large LOCA. Several scenarios of possible sources were identified, and engineering judgment was need om estimate the probabilities of occurrence for the scenarios. A more refined assessment is needed if the indirect scenarios are to become part of the technical basis for decision-making. .

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