ML20209E176

Requests Commission Approval to Publish Notice of Proposed Rulemaking to Modify GDC 4 Re Interim Schedular Exemptions

ML20209E176
Person / Time
Issue date:	03/26/1985
From:	Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
Shared Package
ML20209D524	List: IA-85-409, Recommends Commission Approve Inclusion of Permanent Record Dosimeter in Min Acceptable Personnel Dosimetry Sys for Offsite Emergency Workers Proposed by Federal Radiological Preparedness Coordinating Committee ML20209D521 ML20209D717 ML20209D988 ML20209E149 ML20209E154 ML20209E160 ML20209E165 ML20209E176 ML20209E182 ML20209E389 ML20210F421 ML20210F534 ML20210F656
References
FOIA-85-409, FRN-51FR12502, REF-GTECI-A-02, REF-GTECI-RV, TASK-A-02, TASK-A-2, TASK-OR, TASK-RINV, TASK-SE AB76-2-13, SECY-85-108, NUDOCS 8504050347
Download: ML20209E176 (71)

Text

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%,, ...../ SECY-85-108 RULEMAKING ISSUE (Notation Vote)

For: The Comissioners

(

From: William J. Dircks '

Executive Director for Operations 1

Subject:

PROPOSED' RULE TO MODIFY GENERAL DESIGN CRITERION 4; INTERIM SCHEDULAR EXEMPTIONS

Purpose:

T#obtain Commission approval to publish a notice of proposed rulemaking and staff issuance of schedular exemptions to GDC-4. -

1 Category: This paper covers a major policy issue.

Summary: Advances in technology over the past several years have

led to general acceptance by the NRC staff of the

[. " leak-before-break" concept as applied to large diameter, high-quality piping such as that used in the main coolant loop of pressurized water reactors. This concept would permit, based on detenninistic fracture mechanics and probabilistic analyses, the removal or non-installation of many pipe whip restraints and jet impingement shields originally designed to mitigate the dynamic' effects of postulated instantaneous pipe ruptures. However, GDC-4 l does not permit use of this new technical approach except by exemption. Rulemaking is therefore needed to accomodate this engineering advance, particularly

Contact:

J.A. O'Brien, RES

, 443-7860 -

[

K. Wichman, NRR 492-4679 W. Shields, OELD 492-8693 g l .

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[ The Commissioners -

4 because it has been shown that application of leak ..

before-break to individual plants will result in a safety benefit. Schedular exemptions can'be used on an interim basis pending completion of rulemaking activtties. The staff proposes this rule based on its evaluation of investigations performed by industry and NRC contractors 4'

as well as the staff findings in-the resolution of Unresolved Safety Issue (USI) A-2.

Background:

The regulatory and technical background for this rulemaking is set out in the propcsed Federal Register notice, Enclosure 1. Briefly, the rulemaking stems from the resolution of USI A-2, which considered asymmetric blowdown loads on reactor vessels and their internals and piping caused by postulated instantaneous pipe ruptures

.in:the primary system. Preliminary results indicated that additional restraint structures would be needed to i account for these loads at many operating facilities, entailing considerable utility expense and man-rem expenditure. Westinghouse undertook a generic study to show that primary system piping.would leak at a ,

detectable rate well before a pipe rupture, thus allowing for safe shutdown before a major LOCA could ~ occur. The

results of this study showed this to be the case with high confidence. The staff reviewed and accepted this study as an alternative, basis to resolution of USI'A-2 L set out in Generic Letter 84-04 (Enclosure 2). Research [

performed at Lawrence Livennore National Laboratory

confirmed that pipe ruptures in primary system piping for the three major U.S. vendors of pressurized water reactors were highly improbable.

Because GDC-4 requires that large pipe breaks be

! considered as the design basis, exemptions we m needed to begin regulatory use of the A-2 resolution. To date, two ,

exemptions-have been issued to plants under construction, snd other requests are under staff review. The statf began development of a proposed revision to GDC-4 concurrent with the issuance of Generic Letter 84-04.

Discussion: The Consission's regulations, viz. General Design Criterion 4 and the definition"of "LOCA" used in Part 50,

, Appendix A, require protection against the dynamic effects associated with the postulated ruptures of piping in high energy fluid systems, both inside and outside of containment for structures, systems and components important- to safety' The postulated ruptures include .

i circumferential and longitudinal breaks up to and

~

including double-ended guillotine breaks (DEG8) in the largest pipe in the reactor coolant system.

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The Commissioners )

The need and urgency for addressing the issue stems from the widespread acceptance of the analysis results and the research findings pertaining to pipe rupture coupled with increasing confidence in its applicability. It is now clear that it is possible to defend the exclusion of pressurized water reactor (PWR) primary loop double-ended guillotine pipe ruptures from the design basis and that the scope of this exclusion may be extended to other piping including piping in boiling water reactor:.

Acceptance criteria for generally applying these results pertaining to leak-before-break have been published by the NRC staf f in " Report of the U.S. Nuclear Regulatory Commission Piping Review Committee," NUREG-1061, Volume 3, and are being proposed by the American Nuclear Society in ANS-58.2 entitled " Design Basis for Protection of Light Water Nuclear Power Plants Against Effects of Postulated Pipe Rupture."

General Design Criterion 4, as applied in the context of the definition of "LOCA", has required installation of protective devices (e.g. pipe whip restraints and jet impingement shields) in nuclear power plants to mitigate events wisich are now regarded as extremely unlikely.

These protective devices impede inservice inspection and maintenance, reduce safety if improperly installed, and increase worker radiation exposures.

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c Williad,I o ircks Executive Director for Operations

Enclosures:

1. Federal Register notice

2. Generic letter 84-04

3. Regulatory Analysis 4 Public Announcemes't

5. Congressional letters

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Commissioners' comments or consent should be provided directly to the Office of the Secretary by c.o.b. Wednesday, April 10, 1985.

Commission Staff Offic'e comments, if any, should be submitted to the Commissioners NLT Wednesday, April 3, 1985, with an information copy to the Office of the Secretary. If the paper is of such a nature that-it requires additional time for analytica~1' review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected.

DISTRIBUTION:

Commissioners -

OGC OPE OI OCA OIA OPA REGIONAL OFFICES EDO.

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ACRS ASLBP ASLAP SECY l

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i c NUCLEAR REGULATORY COMMISSION I wasumaron, o.c.zosss N.*****/ February 1,1984 TO ALL OPERATING PWR LI'CENSEES, CONSTRUCTION PERMIT HOLDERS AND APPLICANTS FOR CONSTRUCTION PERMITS

SUBJECT:

SAFETY EVALUATION OF WESTINGHOUSE TOPICAL REPORTS DEALING WITH ELIMINATION OF POSTULATED PIPE BREAKS IN PWR PRIMARY MAIN LOOPS (GENERICLETTER.84-04) i

References:

1. WCAP 9558. Revision 2 (May 1981) " Mechanistic Fracture ~

- Evaluation of Reactor Coolant Pipe Containing a

- Postulated Circumferential Throughwall Crack

2. WCAP9787(May1981)"TensileandToughnessProperties -

of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation"

3. Letter Report NS-EPR-2519. E. P. Rahe to D. G. Eisenhut

' ~

(November 10,1981) Westinghouse Response to Questions and Cossents Raised by Members of ACRS Subcosmittee on

- Metal Components During the Westinghouse Presentation on September 25, 1981.

) The NRC staff has completed its tuview of the above-referenced Westinghouse l '

topical reports and letter report. These reports were submitted to address asymmetric blowdown loads on the PWR primary systems that result from a limited number of discrete break. locations as stipulated in NUREG-0609, the staff's, resolution of Unresolved Saf o Issue A-2.

The staff e. valuation concludes an acceptable technical basis has been provided so that the asymmetric blowdown loads resulting from double ended pipe breaks in main coolant loop piping need not be considered as a design basis for the Westinghouse Owner's Group plants,* provided the following two conditions are met:

1. Reactor primary coolant _ main loop piping at Haddam Neck and Yankee Nuclear Power Station are acceptable provided the results of seismir; analysss confirm that the maximum bending moments do not exceed 42,000 in-kips for the highest stressed vessel nozzle / pipe junction.

• 1. D. C. Cook 1 9. R. E. Ginna

2. D. C Cook 2 10. San Onofre 1

3. H. B. Robinson 2 II. Surry 1

4. Zion 1 12. Surry 2

5. Zion 2 13. Point Beach 1

6. Haddam Neck 14. Point Beach 2

7. Turkey Point 3 15. Yankee .

8. Turkey Point 4 16. FortCalhoun(CENSSS)

Enclosure 1 e4020104g

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2. Leakage detection systems at the facility should be sufficient to provide adequate.marpin to detect the leakage from the postulated circumferential throughwall Mlaw utilizing the guidance of Regulatory Guide 1.45, ,

" Reactor Coolant Pressure Boundary Leakage Detection '

Systems," with the exception that the seismic qualification of the airborne particulate radiation monitor is not necessary. At least one leakage detection system with a t sensitivity capable of detecting 1 gpm in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> must be operable.

j Authorization by NRC to remove or not to install protection against uy m etric -

dynamic loads (e.g., certain pipe whip restraints) in the crimary mata coolant loop will require an exemption from General Design Criteria 4 (GDC-4). , l Licensees must justify such exemptions on a plant-by-plant basis. In such exemption requests, licensees should perform a safety balance in tems of accident risk avoidance attributable to protection from asymetric blowdown loads versus the safety gains resulting from a decision not to use such protection. In the latter category are (1) the avoidance of occupational .

exposures associated with use of and subsequent removal and replacement of pipe whip restraints for inservice inspections, and (2) avoidance of risks associated with improper reinsta11ation. Provided such a balance shows a net safety gain for a particular facility, an exemption to GDC-4 may be -

granted to allow for removal of existing restraints or noninstallation of ,

restraints which would have otherwise been required to accommodate double- -

4 ended break asymetric dynamic loading in the primary coolant loop.

Other PWR licensees or applicants may also request exemptions on the same ,

basis from the requirements of GDC-4 with respect to.asymetric blowdown loads resulting from discrete breaks in the primary main coolant loop, if they can demonstrate the applicability of the modeling and conclusions contained in the referenced reports to their plants or can provide an 4 equivalent fracture mechanics based demonstration of the integrity of the primary main coolant loop in their facilities. -

i t The reports referenced in this letter evaluated the limiting or bounding .

break locations for all the A-2 Westinghouse Owner's Group plants. The  !

l fracture mechanics analyses contained in these reports demonstrated that the potential for a significant failure of the stainless steel primary --

piping was low enough that pipe whip or jet impingement devices for any postulated pipe break locations in the main loop piping should not be ,

i required. The staff's technical evaluation, which is attached, supported the conclusions of the Westinghouse reports. (For infonnation also l attached is the staff's regulatory analysis of this issue.) The staff intends to proceed with rulemaking changes to GDC-4 to permit the use of fracture mechanics to justify not postulating pipe ruptures. The staff will make every effort to expedite rulemaking and will look forward to cooperating with you on this issue.

i i

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,f' t By copy of this generic letter with enclosed. topical report evaluation, and the regulatory analysis, Mr. E. P. Rahe of Westinghouse is being infor1ned of this action.

" incerely, ,

arr G. i'senhut, D tor Division o Licensing Office of Nuclear Reactor Regulati6n

Enclosures:

1. Topical Evaluation Report ,

2. Regulatory Analysis e

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, TOPICAL REPORT EVALUATION Report Title and Number: 1. Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated circum-iferential Throughwall Crack, WCAP 9558,.Rev. 2, Westinghouse Class 2 Proprietary, May,1981.

2. Tensile and Toughness Properties of Primary Piping i Weld Metal For Use In Mechanistic Fracture Evalua-tion WCAP 9787, Westinghouse Class 2 Proprietary, May,1981 l 3. Westinghouse Response to Questions and Comments Raised by Members of ACRS Subcommittee on Metal Components During the Westinghouse Presentation -

._ on September 25, 1981 Letter Report NS-EPR-2519 E.' P. Rahe to Darrell G. Eisenhut, November 10, 1981.

j 1. 0 Background .

L In 1975, the NRC staff was informed of some newly defined asymmetric loads that result by postulating rapid-opening double-ended ruptures of PWR primary piping.

The asymmetric loads produced by ,the postulated breaks result from the theore-tica11y calculated pressure imbalance, both internal and external to the primary system. The internal asymmetric loads result from a rapid decompression that-causes large trarsient pressure differentials across the core barrel and fuel  !

assembly. The external-asymmetric loads result from the rapid pressurization l 3 . of annulus regions, such as the annulus between the reactor vessel and the J,. shield wall, and cause large transient pressure differentials to act on the L

vessel. These large postulated loads are a consequence of the' rapid opening break at the most adverse location in the piping system.

The staff requested, in June 1976, that the owners of operating PWRs evaluate

, their primary systems for these asymmetric loads. Most owners formed owners groups under their respective NSSS vendors to respond to the staff request.

The Babcock and Wilcox (B&W) and Combustion Engineering (CE) owners groups

.each submitted.a probability stiudy, prepared by Science Applications Inc., and

the-Westinghouse owners submitted a proposal for au'gmented inservice inspection.

l The staff reviewed these submittals and concluded at that time that neither l approach was acceptable for resolving this problem. In general, the staff l

' concluded that the existing data base was not adequate to support the con-clusions of the probability study and that the state-of-the-art for inservice inspection' alone was not acceptable for this purpose.

a i

The. staff formalized these conclusions in a letter to the owners of all,operat-ing PWRs in January 1978. This letter also reiterated our desire to'have the PWR owners evaluate their plants for asymmetric loads.. Plant analyses for asymmetric loads were submitted to:the staff for review in March and July 1980. The results of these plant analyses indicated that some plants.would.

r; quire extensive modifications ,1f the rapid-opening double-ended break is rsquired as a design basis postulation.

Also, .in the interim, the technology regarding the potential rupture of rela-tively tough piping such as is used in PWR primary coolant systems, has advanced significantly. Thus, a much better understanding of the behavior-of flawed piping under normal and even excessive loads now exists. The

-NRC staff utilized these technological developments in its review. Tests of deliberately cracked pipes in addition to theoretical fractere mechanics cnalyses indicate that the probability of a full double-ended rupture of tough piping in a typical PWR primary coolant system is vanishingly small.

The subject of PWR pipe cracking is discussed in NUREG-0691 and other references listed in Section 6 of this evaluation.

In parallel with the performance of plant analyses for asymmetric loads, some owners, anticipating potential modifications resulting from the double-ended rupture assumption, engaged Westinghouse to perform a mechanistic fracture evaluation to demonstrete that an assumed double-ended rupture is not a credible design basis ent.i. for PWR primary piping. Upon corpletion of this evaluation, Westinghouse, on the owners group behalf, submitted to the staff for review the topical report, " Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Through-wall Crack," WCAP 9558, Rev. 2. In response to questions raised by the staff, a second report, " Tensile and Toughness Properties of Primary Piping Weld Metal For Use In Mechanistic Fracture Evaluation," WCAP 9787, was also submitted by Westinghouse for our review. In addition, in the

third report listed above, Westinghouse submitted responses to questions and comments of the ACRS Subcommittee on Metal Components during the Westinghouse presentation on September 25, 1981.

2.0 Scope and Summary of Review The analyses contained in WCAP 9558, Revision 2, were performed to demon-strate, on a date inistic basis, that the potential for a significant failure of _ the stunless steel primary piping for the facilities identi- -

fled by the Westinghouse Owners Group was low enough so that main loop pipe breaks need not be considered as a design basis for defining structural loads for resolution of Unresolved Safety Issue (USI) A-2, " Asymmetric Blow-down Loads on Reactor Primary Coolant Systems," or for requiring installation of pipe whip or jet impingement devices for any postulated break location on these lines. Consequently, the staff's review focuses only on the structural integrity of PWR main reactor coolant, loop piaing and does not consider other

t issues such as containment design, release of radioactive materials, or ECCS design at this time. -

Our evaluation includes definition of general criteria that can be used to <

evaluate the inteority of piping.with large postulated loads and cracks.

However, because application of the safety criteria requires system specific input that.would vary significantly in LWR piping systems and because there can be significant differences in pipe loads and materials at various ,other nuclear facilities, our review and conclusions again apply only to the plants named in WCAP 9558, Rev. 2.

Based on our review and evaluation, we have concluded that sufficient technical information has been presented to demonstrate that large margins against

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unstable crack extension exist for stainless steel PWR primary piping postu-lated to have large flaws and subjected to postulated safe shutdown earthquake (SSE)~and other plant loadings. However, several plants in the owners group previously have not performed seismic analyses to define the SSE loading.

These analyses are now being conducted for two domestic facilities as part of the Systematic Evaluation Program. Until the analyses are-completed, we will be unable to make a final decision on the affected facilities. For the remaining facilities included in the Westinghouse Owners Group, the safety margins f indicate that the potential for failure is low enough so that full douole-ended breaks need not be postulated as a design basis for defining st'ructural loads. Also, because the safety adrgins are large, we tentatively conclude that the facilities not having seismic analyses are conditionally acceptable provided that the seismic analyses confirm that SSE' loadings are less than the maximum acceptable levels identified later in this safety evaluation.

The remainder of this safety evaluation includes a summary of the topical reports, ou- evaluation of the reports, and the bases for our conclusions and recommendations. -

3.0 Summary of Topical Reports _ _

l The information contained in topical reports WCAP 9558 Rev. 2, and WCAP 9787 included a definition of the plant-specific primary piping loadings; analyses to define the potentip1 for fracture from ductile rupture and unstable flaw extension; materials tests to define the material tensile and toughness pro-perties; and predictions of leak rate from flaws that are postulated to exist in PWR primary system piping. The essential aspects of these areas are summarized below.

3.1 Loads Reactor coolant pressure boundary (RCPB) piping is required to function under l loads resulting from normal as well as abnormal plant conditions. Loads acting I on the RCP8 piping during various plant conditions include the weight of the l piping and its contents, system pressure; restraint of thermal expansion, i operating transients in addition to startup and shutdown, and postulated i

4 seismic events. In the design of this piping, the limiting loading combina-tion must be determined. The operating facilities that have been evaluated as part of the Westinghouse Owners Group are shown in Table 1.

Based on the loads reported by Westinghouse, bounding loads were defined to envelope the plant-specific. loads; these bounding loads were used in the fracture mechanics analyses that were performed to determine the potential for flaw-induced fracture anywhere within the primary system main loop piping.

3.2 Fracture Mechanics Analysis

, An elastic plastic fracture mechanics analysis was performed to demonstrate that large margins against double-ended pipe break.would be maintained for PWR stainless steel primary piping that contains a large postulated crack and is subjected to large postulated loadings. . Key tasks in the analyses were to determine (1) if the postulated flaw would grow larger oc the application of the load, and (2) if any additional crack growth that might occur would be stable and not result in a complete circumferential break. . The analysis was performed using axial and bending loads that are upper bounds of the loads associated with the facilities identified in Table 1. For aralytical purposes, TABLE 1 Opera' ting Facilities **

Included in Westinghouse A-2 Owners Group Haddam Neck *

] D. C. Cook No.1 & 2 4

R. E. Ginna Point Beach No. 1 & 2

. H. R. Robinson San Onofre No. 1 Surry No. 1 & 2 l Turkey Point No. 3 & 4 i

Yankee Rowe

• Zion No. 1 & 2 Fort Calhoun

" Seismic requirements did not exist for these plants.

• • The Owners Group list of operating facili-ties included a foreign facility, Ringhals No. 2 over which the NRC has no regulatory authority. Thus, we made no formal judgments regarding this facility.

1 a throughwall crack, seven inches in length around the circumference,-war L postulated to exist in the pipe at the section where the bounding bending moments and axial forces occur. This flaw is sufficiently large so that it would be very unlikely to exist undetected during normal operation. (As discussed in NUREG-0691 (Ref. 8), no PWR primary coolant system degradation has been detected to date.)

i The fracture mechanics analysis required determination of a numerical yalue for a parameter that represents the potential for the growth, or extension, of a crack in a pipe that is subjected to specific system loads. This parameter is called the J integral (Ref. 1) and is denoted as J. The J integral is typically employed in fracture evaluations where the section cuntaining the flaw undergoes some plastic deformation due to the loading. Extension or growth of an existing flaw occurs when the value of J reaches a critical value called J initiation, which normally is denoted as Jgg.

When extension of the existing crack is predicted, it is necessary to evaluate this extension and determine if it occurs in a stable manner or if the crack ,

will extend in an uncer. trolled manner and result in a doubled ended break.

The NRC staff requires t.at predicted crack extension be evaluated to assess stability. To comply with this requirement, the Owners Group evaluate'd the predicted crack extension using the tearing stability concept and L% tearing modulus stability criterion (Ref. 2). The tearing stability concept is used i when the mechanism for flaw extension is ductile tearing. This mech d se can I

be expected to prevail for the primary piping materials in the Owners Group's facilities'which are discussed further in the following sections. The tearing modulus is the parameter used to measure the stability of crack extension and

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is denoted as T. Tearing modulus is defined as dJ E - (1)

T=g 7 l

o wherehindicatestheincrementofJneededtoproduceaspecifiedincrement of crack extension at any given load and crack state, l E is the material elastic. modulus, and is the material flow stress defined as one half the sum e, of the material yield and ultimate strengths 6 To determine the margin against fracture, the values.of J and T'are first calculated for the structure using the applied loads and specified crack geometry. The values obtained from the structural analysis create the potential for fracture and are denoted as J applied, or J ,pp, and T applied or T,pp.

l The resistance of the structure to fracture is determined experimentally from l

materials test data that show the relationship between J and crack extension. -

This relationship is called the~ J resistance, or J-R, curve. From this curve the material tearing modulus, or the resistance to unstable crack extension, l is obtained and is denoted as Tmat. At any specified J 1evel greater than-l JIc, stable crack extension wili occur when l' l l

I l

T >

T,pp l mat The amount by which Tmat exceeds T,pp is a measure of the margin against unstable crack extension or, in this case,, the margin against a double-ended break upon application of th'e loading to the flawed pipe.

Topical report WCAP 9558 contains the results of the analyses performed to

. determine J,pp and T,pp. The value of J,pp was determined from an elastic-

, plastic analysis using a finite element computer code. The analysis was based on the bounding load conditions, the postulated seven-inch circumferential throughwall cra:k, and.a lower bound material stress-strain curve obtained at 600*F. The value of T,pp was obtained using previously developed analytical

, methods contained in Reference 3. .

4 The material J-R curves used to determine if crack growth would occur under the postulated loading and flaw conditions and to define values of Tmat 8

defined in VCAP 9558 for base metal and in WCAP 9787 for weld metal. The carbon steel safe-end is discussed in the Westinghouse response to ACRS questions (Subject Document No. 3). A summary of the scope of the materials '

testing follows. .

3.3 Materials Testing Program Base metals representative of those in plants included in the Westinghouse Owners Group were selected for testing. All plants in the Westinghouse Owners Group have wrought stainless steel primary coolant piping except one, which has centrifugally cast stainless steel piping.

Westinghouse selected three heats of cast and three heats of wrought stainless steel for testing. Westinghouse also conducted tests of weld metals to demon-strate that the tensile and fracture toughness properties of the weld metal are comparable to those determined for the base metal in the primary piping system.

A survey of quality assurance files was conducted to identify the primary piping welds in each of the plants in the Owners Group and to define the details of each weld, such as the welding process, electrode size,and material, thermal treatment, and other pertinent information. Based on the survey results, a matrix of representative welding parameters was established and a set of six representative welds was fabricated using typical 2:5-inch-thick base plate.

The welds were then radiographically examined and heat treated where applicable.

Compact tension and tensile specimens were machined from each weld and tested.

Tensile tests were conducted at 600*F using conventional and dynamic loading rates for five of the six heats of base materials. -The sixth heat of base material was tested at conventional loading rates only. Weld metal tensile specimens were tested at conventional loading rates for each weld. Dynamic loading rate tests were not conducted for the weld specimen.

_y-J-resistance (J-R) curves to measure material fracture resistance were generated by multiple specimen testing at 600*F using compact tension specimens at conven-tional and dynamic loading rates for five of the six heats of base metal.

J-resistance curves for the sixth heat of base metal and the weld materials I were generated at 600*F using conventional rates only. The conventional load rate testing and J calculations were perfomed in accordance with the procedures presented in Reference 4. To perform the dynamic toughness test, Westinghousa used a procedure to stop the tests at predetermined displacener.ts, thus allowing development of a J-resistance curve from multiple-specimen dynamic testing.

A minimum of five specimens were tested at conventional and dynamic loading rates for each of the base metal heats. The base metal specimens were machined from pipe sections and oriented so that the crack would grow in the circimferen-tial direction of the pipe. Westinghouse estimated JIe and Tsat **I"'8 I'#

each of the heats of materials tested.

The values of J yg and Tsat were estimated from the slopes of the best-fit straight line through the data points for each base metal' beat. Tut was then adjusted to' account for the nonlinear effects of crack extension using a variation

of the incremental correction scheme suggested by Ernst, et al. (Ref. 5). For the fast rate tests, the data points exhibited a large amount of scatter and, in some cases, there were not enough data points to estinata J yg or Teat

• A minimum of thrse specimens were tested for each weld metal using the same test g procedure f. hat was used for the base metal testing. All of the weld metal data points fell within the scatter band cf the base metal data points except i those for the welds with Inconel filler metal. The data points.for the Inconel weld indicated much higher toughness than any of the other base or weld metals.

Because of the small number of data points, Westinghouse made no attempt at estimating J yg or dJ/da values for the weld metals; however, the weld metal

, data points were fitted with straight lines to demonstrate trends comparable -

to the base metal.

3.4 Leak Rate Calculations To comply with the NRC criteria specified in Section 4.1 for defining postulated flaw size, calculations were performed to define the relationship between leak rate and crack opening area. The leak rate calculations were performed to show that a postul.ated throughwall crack was large enough to produce leaks that could be detected at normal operating conditions by leakage detection devices normally used to detect primary system leakage.

The . leak rate calculations were performed using the method developed by Fauske (Ref. 6) for two phase choked flow; this method was augmented to include frictional effects of the crack surface. An iterative computational scheme was used such that at a given crack operiing area and flow rate the sum of the momentum pressure drop (Ref. 6) and the frictional pressure drop was equal to the pressure drop from the primary system pressure to atmospheric (i.e. ,

2250 - 14.7 psia;.

t To calculate the frictional pressure drop, the relative surface roughness was estimated from fatigue-cracked stainless steel specimens. The leak rate calcula .

tions were performed for a 7-inch-long circumferential throughwall crack at 2250 psi pressure; for conservatism, the bending stress was assumed to be equal to zero for this analysis. The leak rate calculated was approximately 10 gpm.

Although leak rate calculations, especially for small cracks, are subject to uncertainties, the leak rate calculation scheme was correlated with previously generated laboratory data (Ref. 7) and compared with service data from leakage previously detected in the PWR feedwater lines at D. C. Cook and the BWR recircula-tion line at Duane Arnold. In spite of the uncertainties, the calculated leak rate is sufficiently large so as to have a high probability of detection during normal operation. Further discussion of the leak rate analyses is presented in the Westinghouse response to ACRS questions, the third report listed on page one

of this evaluation.

4.0 Evaluation 4.1 NRC Evaluation Criteria The evaluation of the integrity of PWR primary system piping is based on the margin against ductile rupture and resistance to fracture for a postulated throughwall flaw and loading conditions. To determine the potential for flaw- +

induced fracture, the staff required the us, of ar!alysis methods that (1) included an explicit crack tip parameter, (2) predicted the pctential for growth of an existing crack, and (3) determined if any predicted crack exten-sion would' occur in a stable manner. These requirements, coupled with the fact that crack extension in ductile piping material likely will result from ,

ductile tearing, led' the staff to use the J integral based tearing stability '

concept as the basis for our evaluation. The tearing stability concept and-the associated tearing modulus stability criterion (Ref. 2) have been evaluated previously by the staff and found acceptable for use in the evaluation of LWR piping.

The specific criteria used with the tearing stability analysis to evaluate the integrity of PWR primary system piping and determine if adequate margins against flaw-induced failure and pipe rupture are ' maintained include the following:

4.1.1 Loading - The loading consists of the static loads (pressure, deadweight-and thermal) and the loads associated with safe shutdown earthquake (SSE) condi-tions.

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4.1.2 Postu1ated Flaw Size - A large circumferential throughwall flaw is 4 postulated to exist in the pipe wall. The circumferential length of the postulated throughwall flaw is to be the larger of either (1) twice the wall thickness or (2) the flaw length that corresponds to a calculated leak rate

of 10 gallons per minute (gpm) at normal operating conditions.

! Although this safety evaluation has been written exclusively for the primary i system piping at the PWR facilities. listed in Table 1, cracking potential in

. LWR piping is system specific and some additional comments are appropriate concerning the generic application of the assumed flaw sizes used in the piping

9-j '

anslyses. References 8 and 9 indicate that piping systens other than PWR .

primary systems have some service history of observed cracking. For the.se

• systems, consideration should be given to assuming flaw sizes and shapes different from those specified for the PWR primary system depending on the history of observed service cracking, the potential for cracking, and leak detection capabilities. Specific details of LWR piping systems that are sub-ject to cracking, the mechanism for cracking, the nature of the crack sizes and shapes.for these systems, and the effectiveness of flaw and leakage detec-tion methods.are presented in References 8 and 9.

! The NRC staff concludes that the above evaluation criteria are sufficient to I

demonstrate the integrity of PWR primary coolant system piping and that, if met, a break need not be considered anywhere within the main loop piping, thus precluding the need for installation of pipe whip restraints and thus resolving generic safety issue A-2, " Asymmetric Blowdown Loads on PWR Primary

System." As noted in Footnote 1 to Appendix A of CFR Part 50, further details '

l relating to the type, size, and orientation of postulated breaks in specific components of the reactor coolant pressure boundary are under development. We do not anticipate that the final criteria will differ significantly from those ti stated above. Studies and pipe rupture tests have shown that loads far in excess of those specified above still would not result in'a pipe rupture. (These loads might result, for instance, if all the snubbers res1! raining.the steam genera-  !

!. tors were postulated to fail simultaneously. The staff believes this assumption

, to be unrealistic and, if utilized, would depend upon further characterization of material and piping behavior for larger crack extensions.) Other abnormal.

conditions which might affect the' evaluation criteria such as waterhammer,-

stress corrosion cracking or unanticipated cyclic stresses need not be con-sidered for PWR primary coolant main loop piping.

We have reviewed the information provided by Westinghouse relative to the carbon steel safe-ends at the reactor vessel and conclude that our criteria  ;

i also can apply to this piping-to-vessel interface.

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4.1.3 Materials Fracture Toughess Material resistance to fracture should be based on a reasonable estimate of lower bound properties as measured by the materials resistance (J-R) curve.

The lower bound material fracture resistance should be obtained from'either i I archival material of the specific heat of the piping material under evaluation or from at least three heats of material having the same material specification, and thermal and fabrication histories. Both base and weld metal should be tested using a sufficient number of samples to accurately characterize the material J-R curve. To ensure that adequate margins against unstable crack extension exists, the NRC staff concludes that the condition Tmat > 3Tapp should be satisfied at the applied J. level.

4.1.4 Applicability of Analytical Method _

The J-integral and tearing modulus comp'utational methods have certain limits of applicability that are associated with the assumptions and conditions from which they were derived. Generally the limitations are derived from certain

, stress-strain requirements near the crack tip. These requirements translate into restrictions on structural size and material strength and toughness related parameters and are expressed.as (see Refs.-10 and 11) i 1

b > 25 J. (2) and dJ b >> 1 m

= di 3 I3) where b ' = characteristic structural dimension, in this instance pipe wall thickness; -

, a; = material flow stress; -l and dJ = slope of the J-R curve at any given value of J.

di When satisfied, the conditions specified by equations (2) and (3) are suffi-  !

cient to ensure that the J-integral and tearing modulus computational methods can be applied in a rigorous manner and that the results are acceptable for engineering application. The requirement in equation (3) that a >> 1 is some- ,

what indefinite. Generally, a range of a between 5 and 10 satisfies this requirement mathematically and is the range used to perform this evaluation.  ;

While these requirements are used here, they are not necessary conditions.

Less restrictive values (lower value's of b and m) also may be sufficient but l will have to be demonstrated to be so by additional data. These data are not ,

nowavailapleforthepipingmaterialsconsideredinthisinvestigation. .

4.1.5 Net Sectlon Plasticity l The ASME Code specifies margins for pipe stress relative to material yield and-

, ultimate strengths at faulted loading conditions. Because very large flaws may significantly reduce the net load carrying section of the piping, arialyses should be performed to demonstrate that the code limits for faulted conditions are not exceeded for the uncracked section of the flawed piping. Flawed piping having net section stresses that satisfy the code limits for faulted conditions are acceptable. When net section stresses do not meet the code limits, addi-tional analyses or action will be required on a case-by-case basis to ensure that there are adequate margins against net section plastic failure.

4.2 Evaluation Results 4.2.1 Loads -

The loads used to per'orm f the fracture mechanics analyses for the primary piping j include:

I axial tension: 1800 KIPS (includes 2250 psi pressure load), and bending moment:. 45,600 in-XIPS.

These_ loads were derived by " enveloping" the loads obtained from the analyses of record for the highest stressed vessel nozzle / pipe junction of each plant in the Owners Group.

- 11 With the exception of several plants indic'ated in Table 1, the enveloping Joads include those from deadweight, thermal, pressure, and safe shutdown earthquake (SSE) conditions. The static loads (pressure, deadweight, and thermal) were combined algebraically and then summed absolutely with the SSE loads.

The exceptions noted in Table 1 reported axial loads and bending moments that are comprised of only normal operating loads (i.e., thermal, deadweight, and internal pressure) and did not include loads associated with the SSE, the major contributor to the bending moment. Our evaluation is predicated on inciusion of the SSE loadings. However, Connecticut Yankee and Yankee Rowe are being evaluated as part of the Systematic Evaluation Program (SEP) and are committed to perform seismic analyses of their RCPB, safe shutdown systems and engineered safety features using site-specific spectra that will be available in th near future. The completion of such analyses is scheduled for 1983. Confinnation of

! the margins against unstable crack extension under SSE loading will await the seismic analysis of the RCPB main loop piping for these two facilities.

The development of the enveloping loads, including the analytical models, assumptions, and computer codes, were reviewed and approved by the staff

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during'the licensing process for each Owners Group plant and were n'ot reviewed again as part of this effort. We find that these loads, therefore, are upper bound loads and are acceptable for application in the fracture mechanics evaluation of the RCPB main loop piping.

4.2.2 p.teria'1sProperties TensC e Tests - Tensile tests were conducted at conventional and fast loading rates for the base metals and at conventional loading rates for the weld metals.

These tests are relatively straightforward and unambiguous. A comparison of the results from the conventional and fast loading rate tests indicated

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increased yield and ultimate strengths and decreased percentage in elongation at faster loading rates. Except for the weld with the Inconel filler metal, the yield and ultimate tensile strengths for the weld materials were comparable to those for the base metal. The Inconel weld demonstrated a comparable yield i but higher ultimate strength then the base metals. With the exception of the Inconel weld, the percent elongations reported for the weld materials were

. significantly less than those for the base materials, indicating-lower relative ductility for the weldsents.

The tensile properties for the actual base metals in the plants and the test program materials were compatable, indicating that the test materials were representative of the in plant materials. Similarly, the Westinghouse survey of weld materials and techniques was comprehensive and the weld specimens fabricated for testing should be representative of welds in the plants.

Fracture Toughness Testing - Currently, neither an NkC nor a national standard exists for establishing J or J resistance curves, therefore various methods Ic are employed by different laboratories. ^11 fracture toughness testing in the Westinghouse program was performed using the multiple compact tension specimen procedure outlined.in Reference 4.

- 12 This procedure is the basis for the proposed J yg test procedure currently being considered by ASTM Committee E-24 and is generally considered acceptable for determining J The proposed test procedure recommends calculations for deter-Ic.

mining J-Integral values and several criter'ia for ensuring valid J Ic determina-tion. These criteria include considerations of specimen size and data evaluation.

J-Integral Formulation - The expression used by Westinghouse for calculating J for the compact tension specimens has been shown to overestimate the value of J because the experimental data are not corrected for the nonlinear effects of crack growth and plasticity. The effect of this overestimate is to increase

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calculated values of Teat. In rder to account for these effects, Westinghouse applied a correction scheme based on work by Ernst, et al. (Ref. 5). The NRC has reviewed this scheme and found it to be acceptable.

Specimen Size and Geometry - Equations 2 and 3 in Section 4.1.4 specify certain limitations to the applicability of the J-Integral and tearing instability analysis techniques. Because of the high toughness of the heats sampled, not all of the tests satisfied both of these criteria. However, a lower bound .1-R curve, discussed later in this section, was developed for the purpose of this evaluation. This lower bound curve typically meets the requirements of equations 2 and 3 over most of the range of analysis. The exception is for higher levels of J where the specimen dimensions were not adequate as specified by equatior 2. However, the specimen thickness of 1.65 inches to 2 inches for the base me)tals and 2.5 inches for the weld metals approximate the actual thickness of the primary coolant piping (2.5 inches). This similarity in thick-ness simulates the restraint condition in the neighborhood of a crack so that the piping toughness can be represented by the materials test data.

i Side grooving of specimens is a related subject of interest. Side grooving increases the degree of triaxiality in the crack tip stress field and has been shown to result in straighter crack fronts during crack extension. Side grooves are desirable when J-resistance curves are developed using the single specimen -

unloading compliance test or when the data are applied in the evaluation of heavy section structures such as pressure vessels. However, since the specimen dimensions used in these tests approximate the full thickness of the pipes, we conclude that the J-resistance curves developed from specimens without side grooves are acceptable.

Dynamic Tests - The proposed tes' ting procedure used by Westinghouse is intended for quasi-static testing rates. Dynamic toughness t'ests that were conducted l in the Westinghouse program have not previously been performed. Although a full understanding of dynamic fracture toughness in the elastic plastic regime currently is not available, the significant result of the dynamic tests was that the materials consistently demonstrated greater resistance to crack initiation (higher Jyg) at faster loading rates. However, it is noted that two heats of. wrought stainless steel exhibited lower estimated Tsat values at the faster loading rates.

l l

l t

- 13 Based on our review of the materials test data, we conclude that the proposed J-resistance curve test procedure referenced in the subject documents is accept-able for determining J gg and T ,,g. Although the tests conducted did not strictly conform to the criteria recommended in Reference 4, the test specimens -

and procedures are judged to realistically represent the performance of the actual piping systems. In general, the reported ranges of J Ic and T,,g values

!' are acceptable as representative of the structures and materials under consideration.

To perform a generic analysis and account for variations in material behavior, 2 the staff used the data supplied by the Owners Group to define lower bound J-R 1 curves.for the piping materials. The data indicated that Wo lower bound curves l were warranted. One lower bound curve was constructed by a composite of the l wrought and weld data while the second lower bound curve was defined for the cast material. These two lower bound curves were then used with the analyses described in the next section to evaluate the margin against unstable crack extension for wrought and cast stainless steel piping.

l7 4.2.3 Fracture Mechanics Evaluttion l

We have reviewed the elastic plastic fracture mechanics analyses that were  :

submitted by the Owners Group. Our review included independent calculations that were performed to evaluate the acceptability of the Owners Group's

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l conclusiony..

To demonstrate that the postulated throughwall flaw would not sustain unstable crack extension during the postulated loading, finite element calculations first were performed by the Owners Group to determine J,pp as a function of applied.

bending soment with a constant axial force equal to the bounding value of 1800 '

kip's. The relationship between J ,,p and bending soment provided a convenient means to associate the potential for crack extension with the individual plants listed in Table 1.

We have performed independent calculations to verify the relationship between J,pp applied bending acaent. Our calculations are approximate and are based

. on elastic methods corrected for plasticity associated with the loading and the presence of the postulated flaw. While our confirmatory calculations are -

approximations, they do demonstr, ate that the Owners Group calculations are accurate at lower loads where elastic or small-scale, yielding c'onditions prevail G and are conservative at larger loads where plastic deformation occurs. Further, I

the Owners Group elastic plastic analysis is conservative because the analysis was performed essentially for a section of pipe as a free body with applied end loads equal to the bounding loads. This is the limiting (conservative) condition relative to system compliance; a pipe in a real system would be in a less compliant situation and would have lower potential for unstable crack extension.

\

Based on the J ,pp plues calculated for the owners Group by Westinghouse and the' lower bound J-R* curves defined by the staff from the Owners Group materials data, we find that 7 of the 11 United States facilities listed in Table I have sufficient postulated loads to cause extension of the postulated 7-inch-long circumferential throughwall flaw. The loads at the remaining facilities are not high enough to produce extecipn of the postulated flaw.

Of the seven facilities where crack extension was predicted, one has cast

, stainless steel piping. Because of the differences in toughness and tensile properties between the wrought, weld, and cast materials, it was necessary to construct two distinct J-R curves. One curve was constructed from cast material while the second was constructed from a composite of the weld and wrought data.

To determine if the crack extension predicted for the seven facilities would. 1 be stable, the Owners Group'was required to determine the applied tearing modules, T,pp. The value of T,pp was calculated using the methods described in Reference 3.

We have performed independent calculations to verify the Owners Group T,pp calcula-tions using the same methods employed in our J,pp computations. Again, our results '

indicate that the Owners Group calculations are conservative. Based on the obtained from the J-R curve, calculated values of T,pp and the values of Tmat l

we find that large margins against unstable crack extension exist for the seven facilities with predicted crack exte~nsion for the postulated flaw sizes and bending loads.

We also have reviewed the method of analyses that have been performed to estimate the leak rate from the postulated flaw size for nomal operating conditions.

These calculations were perfomed to satisfy a staff requirement that leak detection capability be included, at least cualitatively, in the piping analyses.

Based on our review of the leak rate calculations, we conclude that the calcu-lations presented by the Owners Group represent the state-of-the-art and can be used to qualitatively establish the leak rate for compliance with current

• staff criteria. The leak rate has been determined to be approximately 10 gpm at normal operating conditions and represents, within reasonable limits of accuracy, detectable leakage rates at operating facilities with their available leakage detection systems or devices. For the purposes of this evaluation, there l

i is no need to backfit Regulatory Guide 1.45 to require seismic qualification since such leakage occurs during normal operating conditions.

Based on our review, we have deterstned that all the facilities listed in Table 1.

with the exception of the two facilities 'without seismic analyses, satisfy the acceptance criteria defined.in Section 4.1. Compliance with the acceptance criteria in Section 4.1 ansures that a large margin.against unstable crack i extension exists and that the potential for pioe break in the main loops is suf-ficiently low to preclude using it as a design basis for defining structural loads at the facilities listed in Table 1. .In addition, the facilities that do not have seismic analyses are found to be conditionally acceptable-until the seismic analyses are completed and the loads are defined. Our conditional acceptance is based on: .(1) cur estimate that the seismic loads are not likely to be higher than those listed for the other facilities.in Table 1, (2) the t wide margin against unstable fracture that exists at the maximum moments reported j by Westinghouse, and (3) the los probility that large loadings will occur prior

to completing the seismic analyses.

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Based on our review of the analyses and materials data, we conclude that the-remaining facilities will satisfy all the criteria in Section 4.1 provided that the bending moment in the welded / wrought piping at these facilities does not exceed 42,000 in-kips. If the seismic analyses indicate bending moments in excess of 42,000 in-kips at these two facilities, additional analyses, materials tests, or remedial measures will be necessary to justify these larger

, values. It is noted that the 42,000 in-kip limit applies only to welded / wrought l piping material; a somewhat lower limit would apply for cast material hecause of the. differences in the lower bound J-R curves. However, the facility having the cast material is acceptable and this note is only intended to caution against

the generic use of the 42,000 in-kip limit.

/ The magnitude of-the 42,000 in-kip limit on bending load was determined by find-ing the largest moment that would satisfy the evaluation criteria specified in

, Sections 4.1.3 and 4.1.4 for margin on tearing modulus and size requirements, respectively.

At the 42,000 in-kip load, the margin on tearing modulus.is satisfied.and the value. of m for the test specimens and the primary piping i+; within the specified l range of 5 to 10; however, the value of b for the base metal test specimens is ,

about 30% less than that indicated in equation 2. The lower b value is not a limiting factor in this analysis, however, because as Section 4.2.2 discusses, the specimen thickness is representative of the pipe wall thickness. In addi-tion, the influence of the restriction on size is less than indicated because of the conservatism in the J-integral calculations due to use of a' limiting

compliance condition.

The values of b and a chosen by the staff for our evaluation criteria are '

sufficient conditions and are believed conservative; however, a quantitative estimate of the degree of conservatism cannot be defined without additional experimental data. It is likely that experimental data will shcw that lower values of m and b (and higher allowable moment) could be allowed. Experiments now being conducted or planned by the Office of Research, NRC, and industry organizations such as EPRI should help to clarify this matter in the future.

These additional data are not necessary to complete this review; however, these additional data will be. useful for other studies or for further evaluation of this issue if the bending moments for the remaining facilities are found to i exceed 42,000 in-kips. .

As indicated in Section 4.1, that staff's evaluation criteria are designed to ensure that adequate margins exist against both unst,able fla* e'xtension and net section plasticity of the uncracked pipe section. Both conditions are i evaluated because either may be associated with pipe failure depending on the specific pipe load, material, flaw, and system constraint conditions.

1 Because there may be significant variations or uncertainties associated with these variables, the staff criteria do not attempt to relate margin to actual failure point but is based on maintaining an established margin relative to a

(

combination of conservative bounds for the variables. The margins against actual failure from unstable crack extension are particularly difficult to asstss accurately by analysis because the tough materials used in LWR primary 3

i i

piping typically produce data that fail to satisfy the size restrictions of equations (2) and (3) at the very high J 1evels where failure would be expected to occur.

The 42,000 in-kip limit established by the staff for welded / wrought stainless steel primary PWR piping in Table 1 facilities provides a significant margin against pipe failure. The staff also has reviewed the Owners Group's elastic-plastic analysis and data to provide additional information relative to margin against failure. Based on this review, we conclude that,'for the. conditions cvaluated in this application, the limiting condition is associated with net section plasticity rather ~than unstable crack extension and that the margin -

against net section plastic failure is approximately 2.3 telative to the 42,000 in-kip limit and the postulated 7.5-inch circumfer.ential throughwall flaw. This margin also can be translated into an estimate of margin on flaw size of about 5, i.e., the throughwall flaw size corresponding to net section plastic failure at 42,000 in-kips would be about 38 inches long or 140 degrees around the circumference.

5.0 conclusions and Recommendations

1. Based on our review and evaluation of the analyses submitted for the facilities listed in Table 1, we conclude that the owners Group has shown that large margins against unstable crack extension exist for stainless steel PWR primary main loop pipfr.g postulated to have large flaws and subjected to postulated SSE and other plant loadings. The analytical -

conditions and margins against unstable crack extension satisfy the criteria estaolished by the staff to ensure that the potential for.

L failure is low so that breaks in the main reactor coolant piping up to and including a break equivalent in size to the rupture of the largest pipe need not be postulated as a design basis for defining structural-loads on or-within the reactor vessel and the rest of the reactor coolant system main loops.- Based on compliance with the staff acceptance cri-teria, we conclude that these pipe breaks need not be considered as a design basis to resolve generic safety issue A-2, " Asymmetric Blowdown Loads on PWR. Primary System," for the operating facilities identified a in Table 1. This means that pipe whip restraints and other protective measures against the dynamic effects of a break in the main coolant piping are not required for these facilitiet.

2. Seismic analyses are now being performed for the two domestic facilities listed in Table 1; the reactor primary piping at these facilities are conditionally acceptable and breaks need not.be postulated provided that the seismic analyses confirm that the maximum bending moments do

. not exceed 42,000* in-kips for the highest stressed vessel nozzle / pipe junction.

i

! "For all the facilities listed in Tabla 1, the actual moment is less than 42,000 in-kips and the J,pp is less than J ,,g for each facility.

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3. The criteria used to ensure .tivat ad;qutts, -sargins cgninst brcaks includds .

the potential to tolerate lar@. throughwall flaws without unstable crack extension so.that leakage detaction systems can detect leaks in a. timely -

manner during normal operating conditions. To ensure that adequate leak-detection capability is in place, the following guidance should be satisfied for the. facilities listed in Table 1: .

Leakage detectio>n6ystems should be -stfficient to provide

- adequate maigin toMet ct the leakage from the postulated .

circumferential thro 69twall flaw utilizing the guidance of

Regulatory Guide 1.45, " Reactor.; Coolant Pressure Boundary Leakage'Jetection Systems," with the exceptionithat the seismic qualification of the airtioena particu'.ste radiation monitor is not necessary. At leaston2 leakage' detection system _with a sensitivity capable of detecting 11 gpm in 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. must be operable.

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4. The additional information provided by Westinghouse in response to ACRS questions does not alter cur conclusions. '

6.0 References

1. Rice, J. R. in Fracture Vol. 2. Academic Press. New York,1968

2. Paris,' P. C. , et al. , "A Treatment of the Subject of Tearing Instability,"

U.S. Nuclear Regulatory Commission Report NUREG-0311, August 1977:

3. Tada, H. , et al. .. " Stability Analysis of Circumferential Cracks in Reactor -

Piping Systems," U.S. Nuclear Regulatory Commission Report NUREG/CR-0838, June 1979. -

4. Clarke, G. A., et al., "A Procedure for the Determination of Ductile

• Fracture Toughness values Using J Integral Techniques," Journal of Testing and Evaluation, JETVA, Vol. 7, No.1, January 1979.

5. Ernst, H. A. , et al., " Estimations on J Integral and Tearing Modulus T -

from Single Specimen Test Record," presented at the 13th Material Symposium on Fracture Mechanics, Philadelphia, PA, June 1980.

6. Fauske, H. K., " Critical Two-Phase, Steam Water Flows," Proceeding of the Heat Transfer and Fluid Mechanics Institute, Stanford, California,-Stanford -

University Press, 1961. .

7. Agostinelli, A. and Salamann, V., " Prediction of Flashing Water Flow Through Five Annular Clearances," Trans. ASME, July 1958, pp. 1138-1142.

8. U. S. Nuclear Regulatory Commission, " Investigation and Evaluation of Cracking Incidents in Piping in Pressurized Water Reactors," USNRC Report NUREG-0691, September 1980. ,

9. U. S. Nuclear Regulatory Commission, " Investigation and Evaluation of Stress-Corrosion Cracking in Piping of Light Water Reactor Plants," USNRC Report NUREG-0531, February 1979.

10. Begley, J. A. and Landes, J. D., in Fracture Analysis, ASTM STP 560, American Society for Testing and Materials,1974, pp.170-186.

11. Hutchinson, J. W. and Paris, P. C., " Stability Analysis of J-Controlled Crack Growth," Elastic-Plastic Fracture, ASTM STP 668, American Society for Testing and Materials,1979, pp. 37-64.

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Enclosure 5 Regulatory Analysis of Mechanistic Fracture Evaluation of Reactor Coolant Piping 4

A-2 Westinghouie Owner Group Plants -

P 1.- Statement of the Problem

2. Objective

3. Alternative

4. C'nsequences o .

A. Costs and Benefits I. Introduction II. Values-Public Risk and Occupational Exposure

~~

A. Results B. Major Assumptions III. Impacts-Industry /NRC Costs-Property Damage A. Results B. MajorAssumptions IV. Conclusions B. Impact on Other Requirements l C. Constraints

5. Dec.ision Rationale

6. Im;ilementation

Attachment:

Leak Before Break Value-Impact Analysis

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j Regulatory Analysis of Mechanistic Fracture Evaluation of Reactor Coolant Piping A-2 Westinghouse Owner Group Plants

1. Statement of the Problem The problem of asymmetric blowdown loads on PWR primari systems results from postulated rapid-opening, double-ended guillotine breaks (DEGB) at specific locations of reactor coolant piping. These locations include the reactor pressure vessel (RPV) nozzle pipe interface in the annulus (reactor cavity) between the RPV and the shield wall plus other selected break locations external to the reactor cavity. These postulated ruptures could cause pressure imbalance loads both internal and external to the primary system which could damage primary system equipment supports, core

• cooling equipment or core internals and thus contribute to core melt frequency. ,

l l

This generic PWR issue, initially identified to the staff in 1975, was

~

i designated Unresolved Safety Issue (USI) A-2 and is described in detail l- in NUREG-0609 which provides a pressure load analysis method acceptable to the staff. ,

j The plants to which this analysis applies are the A-2 Westinghouse Owner Group plants identified in Enclosure 2.

2. Obiective _

The objective of this proposed action is to demonstrate that deterministic fracture mechanics analysis which meets the criteria evaluated in Enclosure 2 is an acceptable alternative to (a) postulating a DEGB, j (b) analyzing the structural leads, and (c) installing plant modificatiens l

2-to mitigate the consequences in order to resolve jssue A-2. Demonstrating by acceptable fracture. mechanics analysis that there is a large mar. gin against uristable extension of a crack 'in such piping, (leak before break) contingent upon sati,sfying the staff's. leak detection criteria, will establish a technical justification for the identified plants to be ,

exempted fros-General Design Criterion 4 in regard to the associated ,

definition of a LOCA. Section 4 below provides a Value-Impact assessment of this alternate method for resolving issue A-2 for these plants.

3. Alternative .

The major alternative to the proposed action would be to require each operating PWR to add piping restraints to prevent postulated large pipe ruptures from resulting in full double ended pipe break area, thus reducing the bicwdown asymmetric pressure ' loads and the need to modify equipment supports to withstand those loads as determined in plant specific analysis reported in WCAP-9628 and WCAP-9748, " Westinghouse Owners Group Asymmetric LOCA Loads Evaluation" (Evaluation of DEGB outside and inside the reactor cavity respectively). -

4. Consecuences i

A. Costs and Benefits I. Introduction _

A detailed Value-Impact (V-I) assessment of the proposed alternate resolution of issue A-2 for the 16 Westinghouse A-2 Owners Group

• -'" v- ' p'- "-* "-

plants has been completed by PNL and is attached to this enclosure.

. The V-I assessment uses methods and data suggested in the February 1983 draft of. proposed Handbook for Value-Impact Assessment (PNL4646) and in NUREG/CR-2800, " Guidelines. for Nuclear Power Plant Safety Issue Prioritization Information Development." The nominal estimate results, major assumptions, uncertainties, and conclusions of the assessment are discussed in Sections II, III, and IV below. The results of the upper and lower estimates are incTuded in the table in Section IV below.

II. Values-Public Risk and Occupational Exposure A. Results The estimated reduction in public risk for installing additional pipe res,traints and modifying equipment supports as necessary to mitigate or withstand asymmetric pressure blowdown loads is very small, only about 3 man-rei total for the nominal case for all 16 plants considered. Similarly, the reduction in occupational exposure associated with accident avoidance due to modifying the plants is estimated to total less than 1 man-rem. These small changes result from the e:timated small reduction in core-melt frequency of 1x10-7 events / reactor year that would result from modifying the plants.

However, the occupational exposure estimated for installing

and maintaining the plant modifications would' increase by 1 11,000 man-rem. Consequently, the savings in occupational exposure by not requiring the plant modifications far exceed

~

the potentially small increase in public risk and avoided accident exposure associated with requiring the modifications.

B. Ma_ior Assumotions The above estimated cl.anges in public risk and accident l

avcided occupational exposure were obtained by examining VASH-1400 accident sequences leading to core melt from

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-4 reactor pressure vessel (RPV) rupture and.large LOCA's in conjunction with the major assumptions identified below.

1. If a DEGB occurs inside the reactor cavity, it could displace the RPV, possibly rupturing it or other piping, or disrupt core geometry which could lead directly to core melt in accident sequences analagous to those for RPV '

rupture in WASH-1400.

2. A DEGB in the primary system outside the reactor cavity could lead to core melt through the additional risk contribution from subsequent safety system failures, such as ECCS, induced by previously unanalyzed asymmetric pressure loads on equipment or from core geometry disruptions. It was assumed that failure of safety .

systems independent of asymmetric pressure loading is already accounted for in the plant design.

3. Three sources of data were used to develop estimates of DEGB frequencies, for large primary system piping used in the analysis. These frequency estimates range from an upper estimate of 10-s breaks per reactor year down to a lower estimate of 7x10-12 breaks in a reactor lifetime.

The upper estimate of 10 5/ reactor year is based on a l paper on nuclear and non-nuclear pipe reliability data in IAEA-SM-218/11, dated October 1977 by S. H. Bush which indicates a range of 10

• to 10 5 per reactor year.  !

Additional data in the paper indicates that 10 s may be 100

. times too nigh for the pipe size being considered in issue A-2.

An intermediate or nominal estimate of 4x10-7 per reactor-4 year for primary system piping outside the reactor cavity-and 9x10 5/ reactor year for piping inside the reactor cavity

. . ^*:

• are based on Report SAI-001-PA dated June 1976 prepared-by Science-Applications Inc. which modeled crack propagation in piping subject to fatigue stresses. These values represent an average over p 40 year plant life for a two loop plant and conservatively ignore in-service inspection as a method to discover and repair cracks prior to unstable propagation.

. The lower estimate is based on NUREG/CR-2189, Vol 1, dated September 1981 prepared by LLL. The report uses

, simulation techniques to model crack propagation in primary system piping due 'to thermal, pressure, seismic

. and other cyclic stresses. The report indicates that the probability of a leak is several orders of magnitude more likely than a direct

• seismically induced DEGB which is estimated to have a probability of 7x10 22 over a plant lifetime. For this analysis the lower est'imate of 7x10-12 is considered essentially zero.

It is acknowledged that both the upper and nominal estimate DEGB frequencies used in this analysis are less than the WASH-1400 large LOCA median frequency of 1x10 4/ reactor year. However, the upper estimate of 5/ reactor year is consistent with WASH-1400 median assessment pipe section rupture data. A review of the 16 plants under consideration indicates there are an f

l *Later work (to be published) by LLL indicates that an indirect seismically

incuced DEGB (e.g., earthquake-induced f ailure of a polar crane or heavy "

! component support-steam generator or RC pump) is more probable ranging from

! 10-5 to 10-10/ reactor year with a median of 10-7/ reactor year for plants east

! 'c' the Rockies. Since the nominal DEGB.; frequency obtained from the IAEA paper a::roximates the re:fian indirect DEGB frecuency, the direct DEGB estimate of 7/.0-22 over a plant lifetime was used for the lowewr estimate.

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A average of 10.3 sections of primary system piping per reactor. Multiplying this value by 8.Sx10-f rupture /

section year for large (>3") pipe obtained from Table II' 2-1 results in an estimate of 9x10 rupture / reactor-year. The following table identifies several factors associated with issue A-2' compared to the data base

~

used for WASH 1400 that support use of a lower pipe break frequency

Factor W A-2 Plants WASH-1400 Large LOCA _

Pipe size >30" diameter > 6" diameter Pipe material Austenitic stainless steel Carbon steel and stainless steel System and Class On19 Class I primary system Miscellaneous primary and of pipe pipe with nuclear grade QA secondary system piping' and ISI of various classifications Ty;e of failure Double-endedguillotine(DEG) Circumferential and long-break only itudinal breaks, large cracks Failure location Selected primary system break Random system break locations locations -

L Leak detection LDS capability to detect leak No requirement or provision system (LDS) in a timely manner to maintain for leak detection large margin against unstable crack extension

4. Public dose estimates for the release categories were derived using the CRAC-2 code and assuming the quantities

of radioactive isotopes ~as used in WASH-1400, the meteorology at a tyriical midwestern site (Byron-Braidwood), a uniform population density of 340 pecple per square-mile (which is an average of. all U.S. nuclear power plant sitits) and no evacuation of population. They are based on a 50-mile release r.adius model.

5. The change in occupational exposure associated with accident avoidance assumes 20,000 man res/ core melt to clean up the plant and recover from the accident'as indicated in NUREG/CR-2800, Appendix D.

6. The estimated occupational exposure associated with j installing and maintaining plant modifications considers the plants into two groups. One group of three plants requires extensive modifications according to

~ Westinghouse A-2 Owners Group asymmetric load analysis (WCAP9628). The modifications consisted of added RPV nozzle pipe restraints and substantial modification of all steam generator and pump supports. The occupational exposures for the,se modifications were based on an estimate of 2600 man-rem submitted by San Onofre 1 for modifying three loops. The load analysis for the remaining 13 plants indicates less required plant

modification consisting primarily of RPY nonle pipe

restraints with minor modification of steam generator and/or pump supports for some of the plants. Recalibra-tion of the leak detection systems to assure leak.

detection capability is as'sumed to be required at 14 of the 16 plants and would incur about 200 man-rem total.

III. Imaacts - Industry /NRC Costs - Property Damage A. Results ,

The estimate,d industry costs' to install plant modifications to kithstand asymmetric pressure-loads is about $50 million.

It is, also estimated that power replacane'nt costs would be j an additional $60 million since the plant modifications would be extensive and involve working in areas with limited equipment access and significant radiation levels so that the workl -

would probably extend plant outages beyond normal planned shutdowns. Also, it is estimated that maintenance and inspection of the modifications for the remaining life of all the plants would cost $650K to $1 million in present dollars based on discounting at 10% and 5% respectively. The cost

.for recalibrating leak detection systems is estimated at

about $350K. The above costs do not include the industry costs "

,, .; . expended to date to perform asymmetric pressure load analysis and fracture mechanics analysis. These analyses costs'are ~

considered small compared to the plant mcdifiest hn snd power replacement cost indicated above.

It is estimated that it would cost NRC about 5800K in staff review effort if plant modifications to withstand asymmetric pressure loads were to be installed. If they are not

f . installed and this cost is saved, then it is estimated that NRC cost would be'5400K to review leak detection system calibration work and plant technical specification revisions Exempting the plants from installing modifications would result in a net saving of $400K in NRC costs.

I l . It is. estimated that installing plant modifications to '

withstand asymmetric press'ure loads would avoid public preserty damage costs due to an accident by $24K to 53SK

- -g.

total in present dollar for all the plants based on a discounting at 10% and 5% respectively. Similarly the avoided onsite property damage cost avoided is estimated at.515K to

$29K in present dollars.

Considering the impacts identified.above, it is apparent that the industry and NRC costs savings by not requiring the plant modifications far exceed the small increases in public and onsite property damage costs due to- a potential accident.

B. Major Assumotions _

i 1. The costs for installing the plant modifications were determined by separating the plants int'o two groups.

The cost for the first group of.three plants which

require eitensive modifications used an estimate submitted by San Onofre Unit I which was prorated to the other two plants based on the number of primary loops in each plant. The costs for the remaining 13 plants which would require less modification are derived from Report UCRL-15340 " Costs and Safety Margin of the Effects of Design for Combination of Large LOCA and SSE Leads," and

~

from industry estimates including informal estimates from DC Cook. The estimates were adjusted to 1982 dollars.

2. The cost estiir.ates for public and onsite property damage due to an accident were calculated by multiplying the

~

change in core melt frequency by a generic property damage estimate. This damage estimate was.obtained by using the methods and data in NUREG/CR2723, " Estimates of the Financial Consequences of Nuclear Power Reactor Accidents." Public rIisk upper and lower bound variations are related to Indian Point 2 and Palo Verde values calculated from NUREG/CR 2723.

3. Power replacements costs were based on an assumed 5300K -

per plant outage day.

IV. . Conclusions The resul.ts of the Value-Impact assessment are summarized in the table below. In t'he table, values are those factors relating '

.directly to the NRC role in regulating plant safety, such as reduced public risk or reduced occupational exposure, and are indicated as positive when the results of the proposed action improve p1' ant safety. Impacts are defined as the costs incurred

, as a result of the proposed action and indicated as positive when the resulting costs are increased.

From the table, the main conclusion to be made is that the dose and cost net benefits indicate that not requiring installation of ,

plant modifications to mitigate consequences of asymmetric pressure loads resulting from a possible primary system DEG pipebreak would result in very little increase in public risk and accident avoided occupational exposure (less than 5 man-rem) and ,

would avoid significant plant installation occupational exposure (11,000 man-res) and industry and NRC costs (5110 'million - including 550 million power replacement cost). Three additional observations are worth noting:

a) the uncertainty bounds show net positive benefits for I either dose or cost. The- upperbound is very positive.

b) This assessment does not address costs of core or core support modifications. Adding these costs would increase the avoided cost.

~)

c The cost results are not sensitive to discount rates used in .his assessment.

The detailed PNL Value-Impact assessment is attached to this enclosure.

LEAK BEFORE BREAK VALUE-IM*ACT

SUMMARY

- TOTAL FOR 16 PLANTS Oose (man-rem) Cost (s)

Nominal Lower Upper Nominal Lower Upper Factors - Estimate Estimate Estimate Estimate Estimate Estimate Values (man-rem)

.Public Health -3.4 0 -37 - - -

Occupatjcnal-Exposure -0.8 0 -30 - - -

(Accidental) .

0ccupational Exposure +1.1x104 +3500 +3.2x104 - - -

3perational)

Values Subtotal .+1.1x104 +3500 +3.2x104 - - -

Imoacts (5)

Industry Imgmen-

-50x108 -25x108 -75x108 tation Cost Industry Operating Cost - - -

-6.5x105 -3.3x105 -9.8x105 NRC Development and Implementation Cost 5b)- - - -

-4.0x105 -2.0x105 -6.0x105 Power Replacement Cost - - -

-60x108 -30x108 -90x108 Public Property - - -

+2.4x104 0 +2.6x10' Onsite Property - - -

+1.5x104 0 +4.6x10s Impact Subtotal - - -

-110x108 -55x10' -165x108 (a) Does not include industry costs expended to date to prepare plant asym:netric pressure load analyses and pipe fracture mechanics analysis.

(b) Does not include NRC cost expended to date to develop issue. (NUREG-0609) and to e.aluate Vestinghouse pipe fracture mechanics analysis.

__-__7_________-__.

B. Impact on other Requirements The impact of the proposed action on other requirements is discussed in Section 3.3 of Enclosure.3.

C. Constraints .

Constraints affecting the implementation of the proposed action

. are discussed in Sections 3.5 thru 3.9 and 5.2.1, 5.2.2, and 5.2.3 of Enclosure 3.

5. Decision Rationale _ ,_ ,

The evaluation in Enclosure 2 demonstrates that for the A-2 Westinghouse .0wner Group Plants there is a large margin against unstable crack extension for stainless steel PWR large primary system  ;.

. piping postulated to have large flaws and subjected to postulated SSE

- -- and other plant loads. Having leak detection capability in each of the plants comparable to the guidelines of Regulatory Guide 1.45 (except for seismic I Category air particle radiation monitoring system) assures detecting leaks from throughwall pipe cracks in a timely manner under ,

normal operating conditions; thus maintaining the large margin against unstable crack extension.

- Also, the Value-Impact assessment summarized above indicates that there are definite dose and cost net benefits in not requiring installation of plant modifications to mitigate consequences of a possible primary system p'iping DEG bieak.

6. Imolementation ,

The steps and schedule for implementation of the preposed actior, are discussed in Sections 3.5 thru 3.9 and 5.2.1, 5.2.2, 5.2.3 of Enclesure 3.

. . . _ _ - . - = _ _- . ._ - . - .

i .

LEAK BEFORE BREAK VALUE-IMPACT ANALYSIS i -

1. INT 9000CTI0tl This report presents 'a .value-impact assessment of the consequences of i exempting Westinghouse A-2 Owners Group plants from having to install modifi-

! cations to mitigate asymetric blowdown loads in the primary system.. This . ,

assessment uses methods suggested in the Handbook for Value-impact Assessment. i (Heaberlin et a,1..lgB3) and data developed for, safety issue prioritization 1

( Andrews et al .1983). The assessment relies heavily upon existing industry and NRC reports generated for Generic Task Action Plan (GTAP) A-2, Asymmetric Blowdown Loads on PWR Primary Systems (Hosford 1981).

q

The proposed action will efficiently allocate public resourc

in the generation of electric power and avoid occupational dose wit'h only small .

l increments to public risk. Modification of plant designs to accomodate asymmetric loads in primary systems of selected Westinghouse plants would incur large costs and significant occupational doses for insignificant gains to

'; public safety.

Generic Safety Issue A-2 deals with safety concerns following a postulated major double-ended pipe break in the primary system. Previously unanalyzed loads on primary system components have the potential to alter primary system configurations or damage core cooling equipment and contribute to core melt accidents. For postulated pipe breaks in the cold leg, asymetric pressure

. change's could take place in the annulus between the core barrel and the RPY.

Decompression could take place on the side of the reactor pressure vessel (RPV) annulus nearest the pipe break before the pressure on the opposite side of the RPV changed. This momentary differential pressure across the core barrel induces lateral loads both on the core barrel itself and on the reactor vessel.

Vertical loads are also applied to the core internals and to the vessel because of the vertical flow resistance through the core and asymmetric axial decom-i pression of the vessel. For breaks in RPV nozzles, the annulus between the reactor and biological shield wall could become asymetrically pressurized, resulting in additional horizontal and vertical external loads on the reactor

vessel. In addition, the reactor vessel is loaded simuTtaneously by the j effects of strain-energy release and blowdown thrust at the pipe break. For

' breaks at reactor vessel outlets, the same type of loadings could occur, but '

the internal loads would be predominantly vertical hecause of. the scre-rapid j decompression of the upper plenum. Similar asymnetric forces could also be i generated by postulated pipe breaks located at the steam generator and reactor-coolant pump. The blowdown asymmetric pressure loads have been analyzed and reported in WCAP-9623 (Campbell et al.1980) and WCAP-9748 (Campbell et al.

l Ic79), " Westinghouse Owners Group Asymmetric LOCA Loads Evaluatien."

2.0 PRODOSED ACTIO!: At:0 POTEt!T! AL ALTERr:ATIVES l

l It is prepnsed that Westinghouse A-2 Owner Groue plents listed in j Erc1csu e 2 de exem/.ed from plant modifications to mitipte asyrrnetric blos-l l

1 I

^

down loads to pr arysystemcomphnents. This propssal is based on consider- '

ation of public r'.sk, cccupational dose and cost impacts. The alternative -

would be to require each operating PWR to add pioing restraints and prima,ry system component supports to withstand the blowc;wn asymmetric pressure loads..

Pubite risk reductions for installing / modifying equipment to mitigate <

asymmetric blowdown loads are small. Extensive analyses of pipe material properties and crack propagation by industry (WCAP-9558 and WCAP-9787, Campbell et al,1982 and 1981) and the NRC indicate that catastrophic failures without through-the-wall cracks are extremely 'unlikely. It is proposed that these plants upgrade lesk detection systems, as necessary, to provide adequate leak detection capabilities. This will allow cracks to be identified and repaired before they propagate to. major failures. Plant modifications would increase occupational dose and inspection time for primary -system components. The - i reduction in the frequency of core-Lalt accidents and avoidance of post-i accident doses as a result of the plant modifications is not signif,1 cant.

Cost impacts for equipment to mitigate asymmetric. blowdown loads are plant dependent. In the worst case, they cost many millions pf dollars, require

replacement power purchases and are of questionable feasibility. Some plants considered can handle asymmetric loads with few changes. However, all plants will realize cost savings for the proposed action.

3.0 AFFECTED DECTSION AACTORS _

. Causes Causes

'Ouantified Unquanti fied(a ) g, Decison Factors Change Chanoe Chanoe . .

Public Health X Occupational Exposure (Accidental) X Occupational Exposure (Routine) X Public Property X Onsite Property X Regulatory Efficiency X Inprovements in Knowledge ~~

X Industry Imolementation Cost X Industry Operation Cost X NRC Development Cost X NRC implementation Cost X Np.C Operation Cost X

• ~

d a ; in t'nt s context, "unquantified" means not readily estiidted in'dc11ars.

2 I

A.0 VALUE-IMPACT ASSESSMENT

SUMMARY

- Total for 16 Plants'

,. < Nominal Lower Upper Decisien Factors Estimate Estimate Estimate Values (a) (man-rem)

Public Health -3.4 0 -37 Occupational Exposbre -0.8 n -30

-(Accidental)

Occupational Exposure 1.1E+A 3500 3.2E**

(0perational) .

Regalatory Efficiency N/A Improvements in Knowledge N/A Total Quantified Value 1.1E+4 3500 3.2E+4 Impacts (b) (g)

Industp{) Implementation

Cost \ -1.1E+8 -5.3E+7 -1.6E*8 Industry Operating Cpg 3 -6.5Ev5 -3.3E45 -9.8E+5 NRC Development Cost s '

O 0 -

0

NRC Implementation Cost -4.0E+5 -2.0E+5 6.0E+5

NRC Operation Cost 0 0 0 Public property 2.4E+A n 2.6E-6 Onsite Property 1.5E+4 0 A.6E-5 Total Quantified Impact -1.1E+8 -5.3E+7 -1.6E+8 (a) A decision tern is a value if it supports NRC goals. Principle among these goals is the regulation of safety.

(b) Issacts are defined as the costs incurred as a result .of the proposed action. Negative impacts indicate cost savings.

(c) Does not include industry cost expended to date (fracture mechanics and plant asymmetric pressure load analyses).

Replacement power costs of 56nM are included.

(d) Does not include NRC costs to evaluate asymetric 1 cads (Hosford 1981) or industry fracture mechanics (Campbell 1882).

N/A = Not Affected i

5.0 UNOUANTIFIED *ESIDUAL ASSESSMENT There are no unquantified decision factors in the assessment of this action.

6.0 DEVEL00 MENT OF OUaLIFICATION l A. Public Health a risk analysis was pe-fermed to assess the effects nf exemstir.c l 'r'estinghouse GTA: A-2 ewnar group plar.ts from .odifications to nitijate l

3

.asyernetric blowdcun 1 cads en primary system como:nsnts. This was acccmpitshed .

by exanining WAS

• A00 accident sequences 1cading to core melt from vesspi .,

rupture and large LOCAs.

• For this analysis, it was assumed that a double-ended guillotine (CEG) large LOCA can occur either inside or outside the reactor cavity. In addition to the " standard" stresses caused by a large LOCA (depressurization and loss of coolant inventory), the DEG break can have additional effects:

1. If the DEG break occurs inside the reactor cavity, it can cause an asymmetric blgwdown which displaces the reactor vessel', possibly rupturing other pipes or the vessel itself.

2. If the.0EG . break occurs anywhere in the primary loop, it can cause an asymmetric blowdown which 1) displaces the core such that its geometry i

becomes uncoolable and/or 2) fails needed emergency core cooling system (ECCS) piping through dynamic blowdown forces.

Three sources of data were used to develop estimates of DEG break proba-bilities used in this analysis. These probability estimates range from an upper estimate of IE-5 breaks per reactor year down to a lower estimate of 7E-12 breaks in a reactor lifetime.

The upper estimate is based on a study of nuclear and non-nuclear pipe reliability data (Bush 1977). This data indicates a range of IE A to 1E-6 failures per reactor year. Failures considered include leaks, cracks, ruptures, disruptive and potentially disruptive. Bush indicates values of IE-5 to IE-6 are representative of disruptive failures. A value of IE-5 was used in this analysis as an upper estimate. Additional data presented by Bush indi-cates that this value may be 100 times too high for the pipe sizes being considered in the proposed action.

An intermediate or nominal estimate is based on a study by-SA! (Harris and Fu11 wood 1976) that modeled crack propagation in piping that is subject to fatigue stresses. While the study was done for Combustion Engineering plants, the approach and data are not plant specific. Conservatively ignoring in-

• service inspection as a metnod to discover and repair cracks prior to unstable propagation, SAI reports DEG break frequency estimates of 4E-7/py for the primary system and 9E-8/py in the reactor cavity averaged over a .t0-year plant life for a two loop plant (Figure 23 Harris and Fv11 wood 1976).

The lower estimate of a LOCA was developed by Lawrenc'e Livermore Labor-

. atories (Lu et al.1981) using simulation techniques to model direct effects on crack propagation in primary system pioing due to thermal, pressure, seismic and other cyc1tc stresses. Indirect effects such as external mechanical damage were not included. Results indicate leaks are several orders of magnitude more likely than breaks and that breaks have a probability of 7E-12 over a plant lifetine. This value is essentially zero for risk calculation purposes, so no add,itional lower estimate calculations were performed.

b

It is acknowlaigid that b3th the upper and nominal estinate DEG b'reak frequencies used 7 this analysis are.less than the WASH-1400 large LOCA median l

'. fecauency of IE 4/rsactor-yr. Htwever, the upper cstimato of IE-5/reacter-yeer is consistent with WASH-1400 median assessment pipe section rupture data. A review o.f the 16 plants under consideration indicates there are an average of 10.3 sections of primary system' piping / reactor. Multiplying this value by 8.8E 7 rupture /section-year for large (>1") pipe obtained from Table III .2-1 results in an estimate of 9E-6, ruptures / reactor-year. There are several additional factors associated with this issue compared to the. data used for WASH-1400 that support use .of a lower pipe break frequency. These facters are tabulated below: - -

, Westinghouse A-2 Factor Owners Group Plants WASH-1400 Laroe LCCA

Pipe size - >30 inches diameter - >6 inches diameter Pipe material - austenitic stainless steel - carbon steel and stainless t

steel -

System and class - only class I primary system - miscellaneous primary and of pipe pipe with nuclear grade CA secondary system piping of and ISI varying classification Type of failure - double ended guillotine - circumferential and longitu-(DEG) break only dinal breaks. large cracks Failure location - selected prima'yr system - random system. break break locations locations u

j Leak detection - LDS capability to detect - no requirement or provision system (LDS) leak in a timely manner for leak detection to maintain large margin against unstable crack extensien i

It was assumed that asymmetric blowdown from a DEG large LOCA automatically

causes core melt only if the LOCA occurs within the reactor cavity. Accident

. sequences analogous to those for reactor vessel rupture in WASH-la00 are 4

assumed. These sequences are as follows (Table V.3-14, dominant only):

RC-ci (PWR-11 with frecuency = 2E-12/py

RC-Y.(PWR-2) with frequency = 3E-11/py l RC-6 (PWR-2) with frequency = IE-11/py RC 6 (PWR-2) with f requency = IE-12/py R-c (PWP-3) with frecuency = IE-9/py Rc (PWR-7) with frequency = IE-7/py l WASH-la00 assumes a vesset rupture frecuency of IE-7/py. Replacing this with i 9E-8/py ',the nominal estimate frequency o' in-cavity asymetric blowdown aute-5

-- - - - - - - - . ,_-._.-,m__,-., ,, - __-_.,__-__.,,..---..-,_-.m-. . . - . , . , . _ . . - _ _ - _ _ . . _ . , . _ . - - _ - - - - - _ _ ~ _ . , _ . - - - - _ . . .

matica11y causing core melt in a way analogous to vessel rupture) results in the same orevious equ nce frequencies. , ,

Sese estimates for the release ca egories were derived using the CRAC code

! and Pssur.i g the quantities of radioactive isotopes and guidelines used in WASH-t 140.7, the meteorology at a typical midwestern site (Byron-Braidwood), a uniform populetion density of Oc0 people per square-mile (which is an average of all U.S. nuclear power plant sites)' and no evacuation of population. They are i based or a' 50-mile release radius model.

The nominal estimate risk from the in-cavity DEG large LOCA in a two loop plant becomes:

Risk = (2E-12/py)(5.4E+6 man-rem) + (4E-11/py)(4.8E+6 man-rem) .+

(II-9/py)(5.LE-5 man-rem) + (IE-7/py)(2300 man-ren)

~

= 0.006 man-rem /py I. was assumed that asymmetric blowdown from a DEG large LOCA outside the "

reactor cavity does not automatically Ivad to a core-melt. Subsequent safety systen failures would be needed to result in core-melt, although the potential for the "EG 1arge LOCA to cause such failures directly (or displace the core such that its geometry becomes uncoolable) still exists. ~

Presumably, failure of safety systems independent of asymmetric loading are accounted for in the plant design. Since the DEG break is only part of the WASH-lano large LOCA sequence, it was assumed that no risk is added by the break itself. Only safety system failures induced by unanticipated asymmetric loads on equipment or core geonetry disruptions contribute to this issue.

1 To calculate the contribution to core melt from breaks outside the reactor

,. _ . cavity, a .two-step analysis was followed. First, the contribution to core melt

. fron DIG breeks cutside the reactor cavity was calculated. Second, an -

additionel fraction of this contribution, based on previous systems interaction anllyses was calculated to represent the risk contribution due to asymmetric 4

bloweswe. Only -his fraction would be incurred for the proposed action since DEG reaks were previously considered in the plant design. .

To estir. ate the risk contribution from DEG breaks outside the reactor

. cavity, accicer.t sequences analogous to those for a large LOCA in WASH-la00 are

., assuned applicable. These sequences are as follows (Table V.3-14, dominant l 0-ly } :

l AE-s -(DilR-1} with "

frequency = IE-11/py i AF- c I:bi-1 } " = IE-10/py i

ACD- s (FER-1) = SE-11/py AG- c (?WR-1) = 9E-11/py

17. Y I?UR-2) = IE-10/py i

3. E f

UR-21

= *E-11/:y 4 J u F ~f (:WR-2) = 2E-11/py it.- 2 . ', :ER-?! "

= 2E-R/ry

• M- 2 * : <7 - 2 ) . = IE-6/py i

i 6

i l

i i

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

7

8. NRC OPERATING COSTS It is assened that review of industry sutaittals seeking exclusions from pipe break under the proposed rule change would require the same amount of effort as is currwntly expended reviewing requests for partial exemption from the current version of GDC-4. Therefore, no additional NRC operating costs are anticipated for reactor coolant loop piping. It is anticipated that less effort would actually be required under the. proposed rule change; therefore, a reduction in NRC operating costs could be expected.

s

9. PUBLIC (OFFSITE) PROPERTY DAMAGF' ,

Public property damage. costs were estimated by modifiying the costs from the A-2 evaluation to account for (1) the difference in msnber of plants, (2) the .

difference in average ramaining life of plants, which in turn affects the discount factors applied to costs, and (3) the difference'in core-melt frequency, where applicable. To estimi.te present value, 5% and 10% discount factors of 16.4 and 9.7,~ respectively, were used based on an average of 34.4 years of remaining life.

l These compare to equivalent values in the A-2 evaluation of 13.8 and 9.0, respectively, based on an average remaining life of 23.5 years. The nominal i public property damage costs from the A-2 evaluation were modified as follows:

5% discount factor: Vpp = (.$3. 8 E+4 ) (.85/16 ) (1. 8E-7/1. 0E-7 ) = $4.3E+5 10% discount factor: Vpp = (.$2 . 4 E+ $ } (.85/16 ) ( 1. 8E-7/1. 0E-7 ) = $2.5E+5 The high and low estimates are smynarized below. Note that these incorporate a 150% uncertainty range on the cost per event, plus the 'high and low estimates of core-melt frequency. Note also that no adjustment is made.in the high j estimate of core-melt frequency because it is independent of the number of reactor

! coolant loops.

5% '10%

H.igh Estimate $2.6E+7 $1.5E+7

Low Estbnate 0 0 s

j b

, - - , , -nm - - - . - ,

---,,.-,,g- ,,,,.-.,,-,,,-,,.-,---n.. -

- . , _ _ _ _ - . , . . , - - . . - n, ,,__.___n,n - . - , - ,,,--,-.n---

l 8

10. ONSITE PROPERTY DAMAGE ._

Onsite property damage costs were estimated by modiftying the costs from the A-2 evaluation to account for (1) the difference in the nunber of plants.

(2) the difference in average renaining life of plants, which in turn affects the discount factors applied to costs, and (3) the difference in core-melt frequency, where applicable. To estimate present value, 5% and 'iO% discount factors of 12.9 and 6.1, respectively, were used based on an average of 34.4 years of remaining life. These compare to values of 11.0 and 5.7, respectively, in the A-2 evaluation based on an average remaining life of 23.5 years. The

~ nominal onsite property damage costs from the A-2 evaluation were modified as follows:

5% discount factor: V0P = ($2.9E+4)(.85/16)(12.9/11.0)(1.8E-7/1.0E-7)

$3.3E+5 10% discount factor:

V0P = ( $1. 5 E+4 ).(85/16 ) ( 6.1/ 5 . 7 ) (1. 8 E-7/1. 0E-7$1.5E+5

)

The high and low estimates are stanarized below. Note that these incorporate a 50% uncertainty range on the cost per event, plus the high and low estimates of core-melt frequency. Note also that no adjustment is made in the high estimate of core-melt frequency because it is indpendent of the number of reactor coolant-loops.

5% 10%

High Estimate $d.5E+6 $2.6E+6 Low Estimate 0 0

, 1 EXTENSION OF A-2 VALUE-!MPACT TO ALL PWR PLANTS Garry S. Ilohnan Lawrence Livennore National Laboratory Nuclear Systens Safety Program March 5,1985

i L r .

! 1.' BASIC ASSUMPTIONS The following basic assumptions have been made in extending the usults of l

the value-impact analysis included in NRC Generic Letter 84-04 (" Safe f Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks i in PWR Primary Main Loops").-hereafter referred to as the "A-2 evaluefon", to 1

l plants not addressed by Generic Issue A-2:

i i l l (1) A total of 85 plants are considered, campared to 16 in the A-2 esaluation.

i l

(2) The 85 plants have a total remaining lifetime of 2922 plant-year:,

l compared to 377 plant-years in the A-2 evaluation.

i

(3) The 85 plants have an average remaining lifetime of 34.4 years, capared to 23.5 years for the 16 A-2 plants alone.

I (4) The 85 plants have an average of four reactor coolant loops, com:ared to 3.1 loops in the A-2 plants alone.

i ~

j Basic assumptions made in the A-2 evaluation, as well as sources of c:sts and occupational radiation exposure (ORE) have been retained: .no new sour:ss have' been used, nor have any radical departures been made, for example, ir the

! estimation of public risk.

The estimate of value-impact for all PWR plants based on the A-2 evaluation and these assumptions is sisenarized in the following table. A detailed escription ,

of how these values were arrived at begins in the following section.

i P

i u

2 rO LEAX SEFORI BREAK VALUE-1M*ACT SN - TOTAL FOR 85 PLANTS Dose (man-res) Cost ($)

s Nominal Lower Uppe- Nominal Lower Factors . Estimate Upper Estimate Estimate Estimate Estimate

g. _.

. . _ _ . Estimate Values (man res)  :. -

Public Health -35 0 -3.5E-2 - -

Occupatjonal Exposure (Accidental) -11 0 -2. P 2 - - -

Occupational Exposure p arational) +1.1E+4 +3.5E+3 +3.E-4 - - -

Values Subtotal +1.1E+4 +3.5E+3 +3. E-4 -s, -

Inomets (5)

Industry Ingmen- -

tation Cost - -

-5.0E+7 -2,5E+7 -7.5E+7

. Industry Operating Cost (c) , ,

NRC Development

~

-3.7E+6 -1. 9E+6 -5.6E+6 and Implementation Cost - -

-4.0E+5 -2.0E+5 -6.0E+5 Power Replacement Cost - -

-6.0E+7 -3.0E+7 -9.0E+7 Public Property (c) - - -

+2.5E+5 0 +1.5E+6

. . Dnsite Property (c) - - -

+1.5E45 0 +2.6E+6 Impact subtotal - - -

-114E+6 -57E+6 -154E+6 (a) Does not include ...ww.uy 6vacs expenoen to ca e to prepare plant asyr.ntric pressure load analyses and pipe fra ture mechanics analysis.

(b) Does not include NRC cost expended to date to develop issue (NUREG-0609) and to e.aluate Vestinghouse pipe fracture mechanks analysis. *

(c) Reflects 10% discount factor

(9 1

b

-n~r ,-.,~n,,-w,,._,_,,nn-, ,_,_.--.-_,_?_--.

4 3

c

2. PUBLIC RISK The estimated public risk in the A-2 evaluation was based on WASH-1400

release catagories and a total of 377 plant-years of remaining life for the 16 A-2 plants. Total nominal risk for a two-loop nuclear steam supply system (i.e.,

from both in-cavity and out-of-cavity breaks), based on fracture mechanics results generated by Science Applications Inc., was estimated to be 0.006 man-rem /py.

Assuming that this value can be applied to all 85 PWR plants, and assuming

^

further (and conservatively) that all plants have four loops, the nominal increase in public risk becomes:

Risk = (0.006 man-rem /py)(2922 py)(4/2) = 35 man-rem High Estimate = 351 man-rem Low Estimate = 0 man-rem where the high estimate is based on the 0.12 man-rem /py value developed using the 1E-5/py pipe rupture frequency indicated by Bush that was referenced in the A-2 evaluation. As in the A-2 evaluation, the high estimate is independent of the mnber of loops.

The increase in core-melt frequency developed in the A-2 evaluation is

assumed to be applicable to all plants, modified for the number of loops. The nominal estimate from the A-2 evaluation is:

l l AF = (3.1/2)[9.8E-8/py+0.2(3E-6/py)/250] = 1.4E-7/py Note that in the A-2 evaluation, this value is rounded off to 1.0E-7/py for subsequent calculations of damage costs and accidental ORE. ,

I Assuming that all plants have four loops, the nominal value applicable to l

l all PWR plants becomes: -

AF = (4/3.1)(.1.4E-7/py) = 1.8E-7/py l

l

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

t . 4 '. :. .

High Estimate =

2Ez6/py Low Estimate = 0 where, as in the A-2 evaluation, the high estimate is independent of the number of reactor coolant loops.

3. OCCUPATIONAL EXPOSURE - ACCIDENTAL ... __ ..

The nominal value of ORE due to accidents developed in the A-2 evaluation is modified as follows for differences in total remaining plant life and in  ;

core-melt frequency:

Dose = 0.8 man-rem (1.8E-7/1.0E-7)(2922/377) = 11.2 man-rem High Estimate = 2.3E+2 man-rem. .

Low Estimate = 0 man-rem where the high estimate in the A-2 evaluation has been modified only to account

-for added remaining life (recall that the high estimate of core-melt frequency .

is independent of the number of reactor coolant loops).

l 4. OCCUPATIONAL EXPOSURE - OPERATIONAL . _ -_.... .

Occupational exposure associated with installation of plant modifications

~

applies only to A-2 plants; therefore, the only operating dose estimates from the A-2 evaluation assumed applicable to other plants are those associated with maintenance of the plant modifications (it has been assimed that such maintenance would be typical of that required for PWR pipe whip restraints in general). . Assuming 80 man-hr/py for maintenance activities, the additional avoided operational ORE for plants not include'd in the A-2 evaluation would be:

Dose = (.80 man-hr/py)(.0.025 R/hr)(2922 py - 377 pyl = 64 man-rem

_vge,---.www---r----w-..w, w- -

5

. t The total avoided dose, for all 85 PWR plants therefore becomes:

Dose = Installationdose(A-2 plants) + Operating dose (A-2 plants) +

Operating dose (bther plants)

< = 9700 man-rem ~+ 840 man-rem + 64 man-rem

= 10604 man-rem = 1.1E+4 man-rem i Note that the A-2 value is effectively unchanged by adding the other plants. The high and low estimates are based, respectively, on 3 times and 1/3 times the expected dose as was done in the A-2 evaluation.

4 High Estimate = 3.2E+4 man-rum Low Estimate = 3.5E+3 man-rum

5. INDUSTRY IMPLEMENTATION COSTS Industry implementation costs developed in the A-2 evaluation are for

' installation of specific plant modifications and do not apply to other plants.

Therefore, these costs are assumed to remain unchanged. It has been assumed l that restraint removal in plants not included in the A-2 evaluation would be performed as necessary for in-service inspection or other maintenance. As a r result, the cost of restraint removal would be incurred as a normal cost of maintenance, and not as a specific cost of implementing the proposed rule change.

Note that the A-2 extension does not address avoided costs for other factors affected by the proposed rule change. Elimination of pipe breaks in reactor coolant loop piping would result in (but not necessarily be limited to) reductions'in the costs of:

l

. analysis of thermal hydraulic and piping response to breaks

! - analysis of structural response; structural des.ign design, fabrication, installation, and adjustment of p,ipe whip restraints in plants under construction

- jet impingement target load analysis; design, fabrication, and installation of jet impingement barriers i

e o .

Inclusion of these significant cost items would substantially increase the impact of the proposed rule change. However, they have not been included in accordance with the desire staply to extend the existing A-2 evaluation to other plants. -

4

6. INDUSTRY OPERATING COSTS ,

Avoided costs for operation and maintenance were developed by modifying the l costs from the A-2 evaluation to account for (1) the difference in the number of plants, and (2) the difference in average remaining lifetime of plants, which in ,

turn affects the discount factors applied to costs. The $4540/py cost of cost of l avoided maintenance developed in the A-2 evaluation is assumed applicable to all pWR plants. To estimate present value, 5% and 10% discount factors of 16.3 and 9.6, respectively, are used based on an average of 34.4 years of maaining life. .

These compare to equivalent values of 13.6 and 8.9, respectively, based on an average of 23.5 years of remaining life in the A-2 evaluation. The avoided operation and maintenance costs frcim the A-2 evaluation are modified as follows:

5% discount factor: Cost =

1.0E+6(85/16)(16.3/13.6) = $6.4E+6 ,

10% discount factor: Cost =

$6.5E+5(85/16)(9.6/8.9) = $3.7E+6  : i 5% 10%

High Estimate $9.6E+6 $5.6E+6 Low Estimate . j $3.2E+6 $1.9E+6 where the high and low estimates reflect a 50% uncertainty range.

7. NRC DEVELOPMENT AND I!FLEMENTATION COSTS ,

Development and implementation costs for plants included in the A-2 evaluation have been retained. For other plants, it has been assumed that because an accepta-ble technical basis is available for excluding breaks from reactor coolant loop piping, no additional NRC development effort is necessary. Costs expended during the rulemaking effort have not been included. / l i

i

/

e

. AF. t, (PWR-3\

o IE-8/py AG- 6 (PWR-3. o 9E-9/py ,

Am-6 (PWR-45 = IE-11/py AD. S (PWR-5) "

= AE-9/py AH- ! (PWR-5) "

= 3E-9/py A3 c (PWR-6) = 1E-9/py AHF-c(PWR-6) " *

= IE-10/py ADF-c(PWR-6) "

r 2E-10/py AD- c (PWR-7) ". = 2E-6/py AH. c (PWR-7) = IE-6/py TOTAL 3E-6/py i

WASH-1400 assumes a median large LOCA frequency of IE a/py. Replacing this with 4.0E-7/py (the nominal estimate frequency of outside-of-cavity DEG lary 3 LOCAs) results in lowering the previous sequence frequencies by a factor of 250.- The- risk from the outside-of-cavity DEG large LOCA becomes (ignoring Capendent failures):

i Risk = (IE-12/py)(5.4E+5 man-rem) + (6E-13/py)(4.RE+6 man-rem) + .

(2E-10/py)(5.4E,6 man-rem) + (4E-14/py)(2.7E+6 man-rem) +

(2E-11/py)(1.0E+6 man-rem) + (5E-12/py)(1.5E+5 man-rem) +

(1.2E-8/py)(2300 man-rem)

= IE-3 rian-rem /py

. As assessed in the report for safety issue II.C.3 (Systems Interaction) in Supp. I to NUREG/CR-2800 (Andrews' et al.1983), systems interactions typically contribute 10% to total core-melt frequency (and risk), with a range of 15-20%. The types of safety system failures which could be induced directly by  !

adverse forces from a DEG 1arge LOCA causing asymetric blowdown are typical systems interactions The Westinghouse GTAP *-7 enae s group has provided analyses for ex-cafity breaks that indicate disru:-ic cf core geometry is unitr.ely to occur (Campbell 1980) for 13 out of 16 plants. However, to account for this ;ossibility and

' that of ~ asynmetric-blowdown-induced damage to safety equipment, the uceer end of the range for systems interaction contribution (20%) is assumeo applicable to estimate the risk from dependent failures resulting from outside-of-cavity

, asymetric blowdown. Thus, the incremental best estimate risk from the outside-of-cavity DEG large LOCA with asymmetric loadings becomes:

Risk = (0.2)(1E-3 c.an-rem /py)

= 2E 4 man-rem /py Combining the two scenarios for DEG large LOCAs within and cutside of the reactor cavity yields the following total risk for two loop plants:

Risk = 0.006 + 2E A = 0.006 man-rem /py Nominal estimate results for plants that use a two-loop configuration were adjusted to account fcr the added number:of loops in so a plants. A review cf

~

7

the GTAP A-2 owne s group list indicates that these plants have an average of -

3.1 lotps. The r *.-inal estimate becomes 0.00g man-rem /py. ' " '

Upper estimate risk calculations were made using procedures similar to , ,

those of the nominal estimates. The pipe rupture frequency of IE-5 was allo-cated 8M to the primary loop and 20% to the reactor cavity by assuming the ratio.of results from the SAI study. No corrections for the number of plant '

loops are necessary because this frequency is per plant year. The in-cavity failure rate of 2E-6 is 20 times higher tha'n WASH-la00 for vessel rupture. The upper estimate cavity risk ,becomes:

Risk = ( AE i1[py)(5. AE+6 man-rem) + l A.8E+6 man-rem) +

(8.2E (2.0E-8/py 10/py)h((5 AE+6 man rem) +

(2.0E-6/py)(2300 man-rem) -

l

= 0.12 man-res/py '

The upper estimate of primary loop breaks of 8E-6 'is 12 times lower than i

WASH-1400 for large LOCAs. The upper estimate loop risk becomes: ,

Risk = 0.2 [(2E-11/py)(5.aE+6 man-rem) + (1.3E-11/py)( A.8E+6 man-rem) +

3.9E-9/py)(5. AE+6 man-rem) + (8E-13/py)(2.7E+6 man-rem) +

5.6E-10/py)(1E*6 man-rem

2. AE-7/py)(2300 man-rem)) + (1.0E-10/py)(1.5E+5 man-rem) +

= 0.00A man-rem /py Cembining the two scenarios for" upper estimate break frequencies yields the following total risk:

Risk = 0.12 + 4E-3 = 0.1 man-rem /py Multiplying each of the risk calculations in these cases by the number of-remaining plant years (16 plants x 23.6 yr = 377 py) results in the industry total pubite risk increase due to leak before break.

Total Added Risk (man-rem)

Nominal Estimate 3.A Upper Estimate 37 l Lower Estimate O A nominal estimate for the total increase in core melt frequency for the proposed ec-ion was determined by summing the contribution for breaks inside

• h* reacter cavity and out-of-cavity loop break systems interactions and then ar' justing fer the averep number of looos.

I 8

Core nelt ine , case = 3.1/2[9E-R + 0.2(3E-6/250)1 = IE-7/py

" An upper ostimate of, the ccro-melt frequency increase was calculated by summing the contributions from reactor cavity pipe br.eaks (2E-6/py) and 20",, of -

the out-cf-cavity pipe break initiated core melt accidents.

Core melt increase = 2E-6 + 0.2(2E-7) = 2E-6/py Total core-melt frequency increase estimates are as follows:

Increase in Core-Melt frecuency (Events /oy)'

Nominal Estimate 1E-7 ,

Upper Estimate 2E-6 Lower Estimate 0

8. ' Occupational Exposure - Accidental _

8 The increased occupational exposure from accidents can be estinated as the product of the change in total core-melt frequency and the occupationa,1 exposure likely to occur in the event of a major accident. The change in core melt frequency was estimated as 1E-7 events /yr. The occupational exposure in the event of a major accident has two components. The first is the "immediate" exposure to the personnel onsite during the span of the event and its short term control. The second is the longer term exposure associated with the cleanup and recovery from the accident.

The total avoided occupational exposure is calculated as follows:

DT0A = Nin0AI00A" 'IU10+0LTO)

~

where U 7gg = Total avoiped cccupational dose N = Number of affected facilities T = Average remaining lifetime Pg = Avoided occupational dose per reacter-year D = Change in core-nelt - frequency Djp, = "Imediate" occupational dose Ot7p = Long-term oc:vpational dose.

~

P.esults of the. calculations are shown helow. Uncertainties a e conservative 1v ~

oromagatec by use o' extrews (e.g., uppir bound 073 + upper ocund Cgg).

l l 9 l

t

In . ase in immediateI *I Ltng TernI *) Tctal ' '

Ce*a Melt Occupational Occuoational Avoided Frequency Dose Dose Occupational (events / (man-ren/' (man-rem / Exposure) reactor-yr) event) event)

(man-rem)

. Nominal 1E-7 1E3 2E4 0.A Estimate l

Upper Estimate 2E-6 4E3 3E4 30 Lower Estimate 0 0 1Et 0 i

(a) Based on cleanup and decocraissioning estimates NUREG/CR-2601 (Murphy 1982).

I C. Public Property ,__ . _ _ _

. The effect of the proposed action upon the risk to.offsite property is calculated by multiplying the change in accident frequency by a generic offsite preserty damage estimate. This estimate wak derived from the mean value of _

results of CRAC2 calculations, assuming an SST1 release (major accident), for 154 reactors (Strip 1982). CRAC2 includes costs for evacuation, relocation of displaced. persons, property decontamination, loss of use of contaminated property through interdiction and crop and milk losses. Litigation costs, impacts to areas receiving evacuees and institutional costs are not included.

The damage estimate is converted to present value discounting at 10%. A 5:

discount rate was also considered as a sensitivity case. - - -

The following discounting formula is employed:

• ItI " '-It,*

D=V I

l where D = discounted value V = dar.a ge estinate t .= years before reactor begins operation; O for operating plants i t,g = years remaining until end of life.

i I = discount rate For this proposed action, only operating reactors' are 'affected, and the avet. ige ru=cer of years of remaininc life is 23.5. Therefore, the 107 discount factor D/V = 9. The 5% discount factor equals 12.P., These values must be multiplied by tne number of affected facilities (16): to yield the total effect of the a: tion. Upper and lower beunds are values for Indian Point 2 and Falo Verec 3 calculated from Strip (19R2). Results are as follows:

10

Discounte# Offsite Discounted I-r Proo2rty Damage Value of Additional Of' site Property [ Lifetime Risk) Offsite Property Damage ($/eventi ($/ event) Damage ($)

10% 5% 10% 55 Nominal 't.7E+4 1.5+10 2.3E+10 2.4E+4 3.8E+4 Estimate Upper Estimate 9.2E+9 8.3E+10 1.3E+11 2.6E+6 4.1E+6 Lower Estimate A.3E+S 7.5E+10 1.2E+10 0 0 D. Onsite Property The effect of the proposed action on the risk to onst.te property is estimated by multiplying the change in accident frequency by a generic onsite

. property cost. This generic onsite property cost was taken from Andrews et al. (1983). Costs included are for interdicting or decontaminating onsite property, replacement power and capital cost of damaged plant equipment.

Onsite property damage costs were discounted using the following formula.-

• I D = !1

- (*) .

(1 )

2 f1-e" ") (1-e*II*f**i j 4

where 0 = discounted value V = damage estimate

-m = years over which cleanup is spread = 10 years -

tg = years before reactor begins operation; O for operating plants t, = years remaining until end of life; O = 23.5 years I = discount rate = 10% or 5%.

, For tnis proposed action, the 10% discount factor equals 5.7 and the 5%

discount frctor equals 11. To obtain the total effect of the action, the per-

~

reactor results are evitiplied by the number of affected facilities (16). The uncertainty bounds given in the table reflect a 50% spread which was estimated to be indicative of the uncertainty level. The results are summarized below:

11 l

I

l. _.____ __ _ ___ _ _ _ _ _ _ _ _ . _ _ . . _ . __ _. _ ,

~ .. . . ._

4 Discounted Or 2 *reperty Discount Value of Avoided Damage ! stimate Onsite Property Onsite Property (S/eventi Damage ($/ event) Damage ($1 .

10% 5% 10% 5%

Nominal 1.65E+9 9.aE+9 1.8E+10 1.5E+A 2.9E*4 Estimate Upper Estimate 2.5E+9 .1.4E+10 2.8E*10 4.6E+5 8.8E+5 -

Lower Estimate 8.2E+8 4.7E+9 9.0E+9 0 0 E. Occupational Exposure-operational _ . , _ _

Operational occupational exposure due to installation and maintenance of '

plant modifications is avoided by the proposed exemption to asymmetric blowdown Icads during implementation and operation.

For this analysi,, plants were broken into two groups; those requiring 4

extensive modifications and the rest. . A listing of each group and assumed .

sodifications is given in the section on Industry Implementation Cost. Avoided l implementation dose: for the three plants requiring extensive modifications i

were based on a San Onofre estimate of 2600 man-rem / plant to install primary system pipe restraints at the RPV nozzles and modifying pump and steam generator supports for three loops. Some occupational doses will be incurred for.the proposed action.to upgrade leak detection systems. For these plants, it is estim' ted a that 450 man-hours per plant inside containment at 45 mR/hr and -  !

80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> outside containment at 2.5 mR/hr would be required to install such -

modificaticns. No modifications to the core or core barrel were assumed. For this group, net avoided implementation doses were calculated as follows:

Avoided installation dose = 3[2600 - (0.0025 (80) + 0.045 (450))]

= 7700 man rem Implementation doses for the remaining thirteen plants were estimated as -

follows: 80". of total direct costs were assumed to be attributed to labor in radiation zones. These costs were converted to man-hours by dividing by the cost per man year (assumed to be $100K) and multiplying by 3200 man-hours / man-year. Man-rem estimates were calculated by assnming dose rates of 25 mR/hr inside containment and 2.5 mR/hr outside of containment. The lower value for containrJant work was assumed due to less extensive modifications and presumed better equipment access. Required activities are described further in Industry

'mplementation Costs. - -

4 l

12

Results 'of th': analysis are.as follows:

Humber Dire g of Avoided Ccst Plants D se Rate, Implementation, Activity (S/ loop) (Leops) Man-Hours IbI (R/hr) Dose (man-Rem)

Install primary shield Gall

. restraints and inspection port modifications 98000 13(40)(d.e) 56000 0.025 1400 Modify reactor coolant pump supports 20000 7(21)(d) 6000 0.025 150

~

Steam generator supports 120000 4(12)(d) 21000 0.025 520 Calibrate leakIC) detection system N/A 11(f) 5000 0.025 (120)

. Total 2000 t

(a) Stevenson 1980, except for shield wall and inspection port modifications.

Costs for these-activities are based on industry estimates for 0.C. Ccok.

(b) (Direct Cost)(Number of Loops)(1800 man-hr/ man-yr)(0.8)/(St.0E;5/ man-yr).

(c) Avoided doses are negative for these activities because they are required

'~

for the proposed action.

(d) Campbell 1979 and 1980 (e) Ft. Calhoun was credi'.ed with 3 loops due to redundant cold legs.

(f) Two plants have verified adequate leak detection capability.

Occupational dose to maintain the modifications is also avoided. To estimate the amount, it was assumed that two additional man-weeks per plant-year would De spent inside containment if the modifications are made. This is due to inspection of the modifications and additional time required to gain access to primary system components. The total dose fer the owners groco is esti:.ated below. Plants recuiring extensive modifications have remaining lives totaling 56 plant-years. All other plant lives total 320 plant-years.

Coerational dose averted = (80 a.an-hr/py)[(55 plant-years)(0.r,85 R/ nan ,hr) +

(320 plant-years)(0.025 R/ man-hr)]

= B40 re.an-rem 13

Tctal avside cccupaticnal doses

• due to i plementation, cperaticn and < .

naintenance are nown below. ' Upper and lower estimates were developed using the following model (Andrews et al.1983):

0*S' upper = 3 dose expected Dose ) ,7 = 1/3 dose exhected

~

Activity 00se Avoided (man-rem)- ..

Implemensation 9700 Operation, Maintenance 840 Total 1.1E+4 l- Upper Estimate 3.2E+4 Lower Estimate 3500 i

F. Industry Implementation Cost

.Several levels of yalue to industry are seen as resulting from the proposed action. Potential design modifications that are avoided range from major component support upgrades to the addition of major new equipment, i.e. pipe restraints. Leak detection systems at some plants are already adequate.

Modifications at other plants include an assessment and calibration of existing leak detection systems. The plants were divided into two groups based l on assumed avoided plant modifications:

Plants Requiring Extensive Modifications

. -

Haddam Neck -

Yankee Rowe San Onofre 1 Plants Requiring Some Mcdtfication:

6 Rcbinson 2 -

Zion 1,2 Turkey Point 3,4 RE Ginna Surry 1,2 .

Point Beach 1,2 '

l DC Cook 1,2 Ft. Calhoun.

For plants reoviring extensive modifications, data developed for modifi-l

' cation to prinary system component supports and vessel nozzle restraints by San Onofre-were used (Baskin 1980). Total reported costs were divided by three to cbtain a per-loop cost. Costs for contingencies were ignored. Results are as foli us:

l i

14 l

i

' ~ .

~

• ~

Per-Lotp Costs (SK)

. s Direct Costs (materials field costs) 901 A/E Support -

333

.NSSS Supplier Support 716 Utility Support 166

, Escalation (1979-1982) 740 Total 2856 In addition, Baskin reports that 40 days of replacement power would be purchased. At $300K/ day (Andrews et. al.1983), the total replacement power costs re 512M per plant.

It is conservatively assumed that all three plants will require upgrading

, to their leak detection systems. This may include calibration of current flow measurement systems and revisions to technical specifications. Costs for. these

.; upgrades are based on labor estimates of 0.25 man-yr. At 5100K per man-yr total costs are $25K/ plant.

Total implementation costs for the three plants were calculated as followst Implementation costs ,= (Total Number of Loops)(Avoided Cost per Loop) +

(NumberofAffectedPlants)((ReplacementPower Avoided Cost) - (Leak Detection Costs)3

= (11)($2.85E+6) + 3[SI.2E+7 - 52.5E*43

= $6.7E+7 Implementation costs for the remaining plants are derived from UCP.L-153 A0

! (Stevenson 1980) and industry estimates including San Onofre. Results are indicated below:

\

Mcdification Cost _ _ ~

l. Primary Shield Wall Restraint and Inspection 1230K/ loop Port Modification (Hot and Cold Leg) y Reactor Coolant Pump Suppor,ts S 52K/ loop Steam Geners. tor Supports $310K/ loop Reactor Vessel Supports S 19K/ loop Reactor Coolant Co.ponent

, Walls $230K/ plant The shield wall restra.ints and inspection port modifications are to. control ruptures ir the reactor cavity. These ces;ts were escalated in 1982 dollars based or., estir.ates .fo.- DC Cook units and 'are assumed to include all overheads, i

l 15

,.,.c -

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

material and labe.. All 6ther costs listed are based on work by Stevenscn. .

4 The original wars.cid not appear to inclu'de enginsering, NSSS supplier a'ndr * .

utili y support costs. An additional 13a: was assumed for these costs based on tne San Onofre data. All costs were also increased by an additional 19% for '

escalations between 1980 and 1982. i All modifi. cations would net be required at all plants. Based on Owners -

l Group analyses (Campbell 1979), it was assumed that the following number of l modifications would be performed.

Dwners Group Avoided Modification Number of Plants (Looos) Cost _

Primary Shield Wall 13 (40) $9200K-Restraint and Inspection -

Port Modification Reactor Coolant Pump 7 (21) $1100K Supports Stean Generater Supports 4 (12) 537DOK Reacter Vessel Supports 0 0 Reacter Coolant Conpartment 0 -

0 Walls ,

Total ~ ~ $1A000K Shield wall restraints and inspection port modifications were assumed to be '

required at~all plants. Pump and steam generator support work was assumed to be needed'at plants identified by the owners group. Reactor vessel supports were assumed not to be needed by any plants. Stevenson-discusses them as mainly'a seismic res,traint. Reactor coolant compartment wall anchors are only required for the safe shutdown earthquake (SSE) and LOCA load combinations.

Thus they were not used in this analysis.

Needs for replacement power to modify remaining plants were not identified in the available data. It was assumed for plants requiring pump and steam generator support modifications that soae replacement power would be needed ,

(four plants). For this analysis, it was assumed that one half of the

~

increr. ental outage time of San Onofre would be needed or 20 days. Total outage days would be 80. Costs for replacement power at S300K/ day total 524M. -

Cests for modifying leak detectio.t systems are assumed the same for plants recuiring some modification'as for plants with extensive modificatiens. It was assumed that only 11 of the 13 plants need upgrading. Costs for this work total 52.RE*5.

!' ;et avoided ccsts for plants with some modifications were calculated as failews: ,

I 16 t-

. - r Wet Avsided Implementation Costs *= Primary System Modificatices + -

Eeplacenent Power - Leskaga Detection

.,.' , Systems.

= $1.dE+7 + S2.4E+7 - 52.85+5

~

= $3.RE+7 To gene' rate. upper and lower estimates for costs, it was assumed that esti-mates are within 50? of the nominal estimate. Results' for industry implemen-tation corts are sunmarized below:

Plants with Extensive Modifications $6.7E+7 Plants with Some Modifications $'3.8E+7 .

. Total $1.1E+8 Upper Estimate $1.6E+8 Lower Estimate $5.3E+7 G. Industry Operation and Maintenance Costs _

Industry ec.*ded operation and maintenance costs were. developed based on the assumption that Mdditional restra.ints will result in additional inspections and restrict access to steam generators, reactor coolant pumps and reactor i nozzles. Based on the values used for occupational dose estimates, this labor is assumed to total 80 man-hours / plant-year. At 5100K/ man-year and 44 man-wk/ man-yr, the annual cost is $4540/ plant. The present value of this quantity for 16 plants over 23.5 years with upper ar.d lower estimates are as .follows:

Discount Rate 10* si Present Value of Operation

. and Maintenance Costs = 56.5E*5 1.0E*6 Upper Estimate = $9.8E*5 1.5E+6 Lower Estimate = $3.3E+5 5.0E+5 H. MRC Implementation Suerort rests NRC Avoided implementation costs are estimated to be 0.5 man-year of labor to review plant modifications. This is partially offset by an estima e of 0.25 nan-year to review leak detection system upgrades and revisions to plant technical specifications. Net NP.C cost savings ere as follows: '

17

. - . e Avaided NRC Ino1ementation Supp3rt Costs: .

15 plants (0 25 man-yr/ plant 6 5100,000/ man yr) = $4.0E+5 Upper Estiante = $6.0E+5 -

Lower Estimate = $2.0E+5 ,

No additional NRC costs during operations Are expected.

7.0 CCNCLUSIONS --.

The summary results for the value-impact assessment are shown 'oelow. The nominal estimates for cost and dose indicate that the proposed action should be .

recommended. The uncertainty bounds .do not show negative. benefits for either dose or tost. The upper estimate is very positive. The following observations can also be mark:

o This action did not address costs-of core and core support. modifications.

Adding these costs would increase the negative impact of the exemption.

s The schedule for avoided plant modifications assumed backfitting to add only an increment of downtime to normal outages. If not, the additional avoided costs for replacement power would increase the negative impact -

obtained, o The dose avoided for this action is primarity occupational dose during equipment installation. This dose is being weighed against statistical estimates of public and cccupational dose for rare events. .

o Cost results are not sensitive to discount rates used in this analysis.

Sumaav of Value-Imoact Assessment __i,_ _

l Value Inea ren1 inoact 'S)

Nominal Upper Lower Est. Est. Est. Nominal Est. _

Uoper Est. Lower Est. _ _ . , .

10% 5% 10". 55 10" 5% _

1.'E*4 -3.2E-4 35nn -1.1E+a -1.1E*8 -1.6E+8 -1.6E 8 -5.3E+7 -5.3E*7 l

l 18 l 1

I

_ _ _ _ . _ . _ . - - . - _ - _ . - - _ _ . _ = _ . . . _ _ _ _ - - _ _ - . . . _ - - , - . _ - __, . - - _ _ . . _ _ - - -

" ' " . REFERENCES f .' ildrich, D. C., et a1. 1982. Technical' Guidance for Siting Criteria D:velcoment. NUREG/CR-2239, Sanoia National Laooratories,'Alouquerque, New Mexico.

Anciews, W. B., et al. 1983. Guidelines for Nuclear Power Plant Safety Issue

,Prioritization Information Develooment. NUREG/CR-2800 (PNL 4297), Pacific hor nwest Lacoratnry, Ricnlano,"Wasnington.

Baskin, K. P. 193.0. Letter to Mr. D. L. Zuemann of the US NRC dated Februa ry 13, 1980 Docket No. 50-206. Souther'n California Edison, P.osenmad, California.

Bush, S. H. 1977.* " Reliability of Piping in Light Water Reactors." IAEA-SH-21R/11. International . Symposium on Application of Reliability Technology to

} Nuclear Power Plants. International Atomic Energy Agency, Vienna, Austria.

Campbell , T. E. et al . 1980. Westinghouse Owners Group Asymetric LOCA Loads

. Evaluation Phase C. WCAP 97a8, Westingnouse Electric Corp., Pittsourgn,

, Pennsylvania.

$ Campbell , T. E. et al . 1079. Westinghouse Owners Group Asymetric LOCA Loads Evaluation Dhase 3. WCAP 9628, Westingiouse Electric Ccrp., Pittsourgn, h Fennsyl vania .

3 Campbell, T. E. and J. N. Chirigos , et al . 1882. Mechanistic Fracture Evaluation of Reactor Coolant Pipe .Containing a Postulated inecugn-W'all cract. WCI.P 955 , P.ev. 2, Class 2, Westingnouse Electric Corp., Pittsourgh, -

g Fennsyl vania .

B Camobell, T. E. and J. N. Chirigos, et al. 1981. Tensile Toughness Procerties of Primary Piping Weld Metal for Use Mechanistic rracture ~ ~ ~

Eva 4uation. WCeP 978i, Class 2 Westingnouse Electric Corp., Pit:sourgh,

~ ~ ~ ~ --^

Pennsylvania.

• 6 i Harris, D. O. and Fu11 wood,.R. R. 1976. An Analysis of the Relative Peebability of Pice Ruoture at Various locations in tne Primary Cociing-Loco of a ? essurizec water ?.eacter Inclucing ine Effects of Pericc1c Ins:ect'on.

l 5Al-Gul-PA, Science Applications Inc., Palo Alto, California.

l. Heaber.lin, S. W., et al. 19*3. A Fandbook for Value-teop.ct *ssesseeet.

~ ~ ~

o*:L 454E (Draft), Pacific Northwest Lacoratory, Richlano, hasmng :n.

l

~

Hosferd, L. B.. et al. 1981. Asymetric Blowdown loads en Pk?. F-imary

l. . . Systees. -NUREG 06C9 U.S. Nuclear Regula ory Cormission, .casr.ing en. D.C.

.u, S. 1981. : obability of Pipe Fracture in the Primary Cc.cler.: Loro of a O G. D 1 7. r t. . MUPistCR-21e9. U.S. Nuciea r f.egula: cry Co=1ss1:5, r'.ksning ,on ,

l

.urpry E.-S., and G. M. Halter. 1982. Tachnology, Safety and Costs ' 's 1I o' Dsec~.issicnir Deferonce Light Water Reactors Following Postulated ,

'cc:r.er s. tiuREG,a-zent, asciric Nor n.ast Laooratory, Ricniano, wa s ni n gton. (

Stevenson, J. D. '1980,. Cost and Safety F*rgin Assessment cf the Effects of s Design for Combination of Larne LOCA and SSE Loaos. UGL-15340, Lawrence -

Liver:nore Laooratory, Livermore, Calif ornia.

Strip. D. R. '1982. Estimates of *the Financial Consecuences of Nuclear Power _.

Reactor Accidents. tWEG/ER-2723, Sanoia National Laboratories,- Albuquerque, New Pexico. -

4 i

i

• , s, 10

... ,mm..m..im e nn

,---------w, - , , . , , _ _ , , _ , _ _ _ _ . - _ , _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ . - _ _ . . - . - _ , _ _ , _ , _ _ . -_ _ . ,_,_ _