NRC Generic Letter 1984-04
| ML031150562 | |
| Person / Time | |
|---|---|
| Issue date: | 02/01/1984 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| GL-84-004, NUDOCS 8402010410 | |
| Download: ML031150562 (55) | |
I UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, 0. C. 20555 February 1, 1984 HOLDERS AND
PWR LICENSEES, CONSTRUCTION PERMIT
TO ALL OPERATING PERMITS
APPLICANTS FOR CONSTRUCTION
REPORTS DEALING WITH
SAFETY EVALUATION OF WESTINGHOUSE TOPICAL
SUBJECT: BREAKS IN PWR PRIMARY MAIN LOOPS
ELIMINATION OF POSTULATED PIPE
(GENERIC LETTER 84-04)
"Mechanistic Fracture
1. WCAP 9558, Revision 2 (May 1981) Pipe Containing a References:
Evaluation of Reactor Coolant Throughwall Crack"
Postulated CircumferentiaL
Properties
9787 (May 1981) "Tensile and Toughness Mechanistic
2. WCAP for Use in of Primary Piping Weld Metal Fracture Evaluation"
9 , E. P. Rahe to D. G. Eisenhut
3. Letter Report NS-EPR-251 Response to Questions (November 10, 1981) Westinghouseof ACRS Subcommittee on Members and Comments Raised by the Westinghouse Presentation Metal Components During on September 25, 1981.
Westinghouse staff has completed its review of the above-referenced submitted to address The NRC report. These reports were a
topical reports and letter on the PWR primary systems that result from the asymmetric blowdown loads break locations as stipulated in NUREG-0609, limited number of discrete Safety Issue A-2.
staff's resolution of Unresolved has been provided concludes an acceptable technical basis ended pipe breaks The staff evaluation blowdown loads resulting from double for the so that the asymmetric piping need not be considered as a design basis conditions in main coolant loop plants,* provided the following two Westinghouse Owner's Group are met: Neck primary coolant main loop piping at Haddam provided
1. Reactor Station are acceptable and Yankee Nuclear Poweranalyses confirm that the maximum the results of seismic exceed 42,000 in-kips for the highest bending moments do not junction.
stressed vessel nozzle/pipe
9. R. E. Ginna
1. - D. C. Cook 1 10. San Onofre 1
2. D. C Cook 2 11. Surry 1
3. H. B. Robinson 2 12. Surry 2
4. Zion 1 13. Point Beach 1
5. Zion 2 14. Point Beach 2
6. Haddam Neck 15. Yankee
7. Turkey Point 3 16. Fort Calhoun (CE NSSS)
8. Turkey Point 4 Enclosure 1
< O 41
] 0
-2
2. Leakage detection systems at sufficient to provide adequatethe facility should be leakage from the Postulated margin to detect the flaw utilizing the guidance of circumferential throughwall
"Reactor Coolant Pressure Boundary Regulatory Guide 1.45, Systems," with the exception Leakage Detection of the airborne particulate that the seismic qualification necessary. At least one leakage radiation monitor is not sensitivity capable of detecting detection system with a operable. 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 Authorization by NRC to remove dynamic loads (e.g., certain or not to install protection pipe against asymmetric loop will require an exemption whip restraints) in the primary from General Design Criteria main coolant Licensees must justify such 4 (GDC-4).
exemption requests, licensees exemptions on a plant-by-plant basis.
In should accident risk avoidance attributable perform a safety balance in terms such of loads versus the safety gains to protection from asymmetric resulting from a decision not blowdown protection. In the latter category to use such exposures associated with use are (1) the avoidance of occupational pipe whip restraints for inserviceof and subsequent removal and replacement of associated with improper reinstallation. inspections, and (2) avoidance of risks net safety gain for a particular Provided such a balance shows facility, an exemption to GDC-4 a granted to allow for removal may be restraints which would have of existing restraints or noninstallation otherwise been of ended break asymmetric dynamic required to accommodate double- loading in the primary coolant loop.
Other PWR licensees or applicants basis from the requirements may also request exemptions of GDC-4 with respect to asymmetricon the same loads resulting from discrete breaks in the primary main coolant blowdown if they can demonstrate the applicability loop, contained in the referenced of the modeling and conclusions equivalent fracture mechanicsreports to their plants or can provide an primary main coolant loop in based demonstration of the integrity of the their facilities.
The reports referenced in this break locations for all the letter evaluated the limiting A-2 Westinghouse Owner's Group plants. or bounding fracture mechanics analyses contained in these reports demonstrated The the potential for a significant that piping was low enough that pipe failure of the stainless steel primary postulated pipe break locations whip or jet impingement devices for any required. The staff's technical in the main loop piping should not be the conclusions of the Westinghouse evaluation, which is attached, supported attached is the staff's regulatory reports. (For information also intends to proceed with rulemaking analysis of this issue.) The staff fracture mechanics to justify changes to GDC-4 to permit the not postulating use of will make every effort to expedite pipe ruptures. The staff cooperating with you on this rulemaking and will look forward issue. to
-3 -
I .
evaluation, of this generic letter with enclosed topical report is being By copy Mr. E. P. Rahe of Westinghouse and the regulatory analysis, informed of this action.
sincerely, abr G. isenhut, irector Division o' Licensing Office of Nuclear Reactor Regulation Enclosures:
1. Topical Evaluati n Report
2. Regulatory Anal sis
TOPICAL REPORT EVALUATION
Report Title and Number: 1. Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circum- 2, ferential Throughwall Crack, WCAP 9558,-Rev.
Westinghouse Class 2 Proprietary, May, 1981.
Primary Piping
2. Tensile and Toughness Properties of Evalua- Weld Metal For Use In Mechanistic Fracture tion, WCAP 9787, Westinghouse Class 2 Proprietary, May, 198.1.
Comments
3. Westinghouse Response to Questions and on Metal Raised by Members of ACRS Subcommittee Components During the Westinghouse Presentation on September 25, 1981, Letter Report NS-EPR-2519,
10,
E. P. Rahe to Darrell G. Eisenhut, November
1981.
1.0 Background newly defined asymmetric loads that In 1975, the NRC staff was informed of some ruptures of PWR primary piping.
result by postulating rapid-opening double-ended breaks result from the theore- The asymmetric loads produced by the postulatedinternal and external to the primary tically calculated pressure imbalance, both from a rapid decompression that system. The internal asymmetric loads result across the core barrel and fuel causes large transient pressure differentials result from the rapid pressurization assembly. The external asymmetric loads the reactor vessel and the of annulus regions, such as the annulus between differentials to act on the shield wall, and cause large transient pressure a consequence of the rapid-opening vessel. These large postulated loads are piping system.
break at the most adverse location in the owners of operating PWRs evaluate The staff requested, in June 1976, that the loads. Most owners formed owners their primary systems for these asymmetric to respond to the staff request.
groups under their respective NSSS vendors Engineering (CE) owners groups The Babcock and Wilcox (B&W) and Combustion by Science Applications Inc., and each submitted a probability study, prepared for augmented inservice inspection.
the Westinghouse owners submitted a proposal at that time that neither The staff reviewed these submittals and concluded problem. In general, the staff approach was acceptable for resolving this not adequate to support the con- concluded that the existing data base was the state-of-the-art for inservice clusions of the probability study and that purpose.
inspection alone was not acceptable for -this
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The staff formalized these conclusions ing PWRs in January 1978. This in a letter to the owners of all PWR owners evaluate their plants letter also reiterated our desire to haveoperat- the asymmetric loads were submitted for asymmetric loads. Plant analyses for
1980. The results of these to the staff for review in March plant analyses indicated that some and July require extensive modifications plants would if the rapid-opening double-ended required as a design basis postulation. break is Also, in the interim, the technology tively tough piping such as regarding the potential rupture-of is used in PWR primary coolant systems, rela- advanced significantly. Thus, has of flawed piping under normal a much better understanding of the behavior and even excessive loads now NRC staff utilized these technological exists.
of deliberately cracked pipes developments in its review. The in addition to theoretical fracture Tests analyses indicate that the probability mechanics tough piping in a typical PWR of a full double-ended rupture primary coolant of The subject of PWR pipe cracking system is vanishingly small.
references listed in Section is discussed in NUREG-0691 and
6 of this evaluation. other In parallel with the performance owners, anticipating potential of plant analyses for asymmetric modifications loads, some rupture assumption, engaged resulting from the double-ended evaluation to demonstrate thatWestinghouse to perform a mechanistic an assumed double-ended rupture fracture credible design basis event is not a this evaluation, Westinghouse, for PWR primary piping. Upon completion of the staff for review the topicalon the owners group behalf, submitted to of Reactor Coolant Pipe Containingreport, "Mechanistic Fracture Evaluation wall Crack," WCAP 9558, Rev. a Postulated Circumferential staff, a second report, "Tensile2. In response to questions raised Through- and Toughness Properties of by the Piping Weld Metal For Use In Primary Mechanistic Fracture Evaluation,"
was also submitted by Westinghouse third report listed above, for our review. In addition,WCAP 9787, Westinghouse submitted responses in the and comments of the ACRS Subcommittee to questions Westinghouse presentation on on Metal Components during the September 25, 1981.
2.0 Scope and Summaryof Review The analyses contained in WCAP
strate, on a Jete inistic basis, 9558, Revision 2, were performed to demon- that the potential for a significant failure of the stainless steel fied by the Westinghouse Owners primary piping for the facilities identi- pipe breaks need not be considered Group was low enough so that main loop loads for resolution of Unresolved as a design basis for defining structural down Loads on Reactor Primary Safety Issue (USI) A-2, "Asymmetric Coolant Systems," or for requiring installationBlow- of pipe whip or jet impingement these lines. Consequently, devices for any postulated break the staff's review focuses only location on integrity of PWR main reactor on the structural coolant loop piping and does not consider other
- 3 -
materials, or ECCS
issues such as containment design, release of radioactive design at this time.
criteria that can be used to Our evaluation includes definition of general postulated loads and cracks.
evaluate the integrity of piping with large criteria requires system specific However, because application of the safety piping systems and because there input that would vary significantly in LWR loads and materials at various other can be significant differences in pipe again apply only to the nuclear facilities, our review and conclusions plants named in WCAP 9558, Rev. 2.
have concluded that sufficient technical Based on our review and evaluation, we that large margins against information has been presented to demonstrate steel PWR primary piping postu- unstable crack extension exist for stainlessto postulated safe shutdown earthquake lated to have large flaws and subjected several plants in the owners group (SSE) and other plant loadings. However, to define the SSE loading.
previously have not performed seismic analyses two domestic facilities as part of These analyses are now being conducted for be the Systematic Evaluation Program. theUntil the analyses are completed, we will unable to make a final decision on affected facilities. For the remaining Owners Group, the safety margins facilities included in the Westinghouse is low enough so that full double- indicate that the potential for failure a design basis for defining structural ended breaks need not be postulated as are large, we tentatively conclude loads. Also, because the safety margins analyses are conditionally acceptable that the facilities not having seismic that SSE loadings are less than the provided that the seismic analyses confirm in this safety evaluation.
maximum acceptable levels identified later includes a summary of the topical The remainder of this safety evaluation and the bases for our conclusions and reports, our evaluation of the reports, recommendations.
3.0 Summary of Topical Reports WCAP 9558, Rev. 2, and WCAP 9787 The information contained in topical reports primary piping loadings; analyses included a definition of the plant-specificductile rupture and unstable flaw to define the potential for fracture from material tensile and toughness pro- extension; materials tests to define the flaws that are postulated to exist perties; and predictions of leak rate from aspects of these areas are in PWR primary system piping. The essential summarized below.
3.1 Loads piping is required to function under Reactor coolant pressure boundary (RCPB) plant conditions. Loads acting loads resulting from normal as well as abnormal include the weight of the on the RCPB piping during various plant conditions restraint of thermal expansion, piping and its contents, system pressure; and shutdown, and postulated operating transients in addition to startup
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,
,
seismic events. In the design tion must be determined. of this piping, the limiting The operating loading combina- part of the Westinghouse Owners facilities that have been evaluated Group are shown in Table 1. as Based on the loads reported envelope the plant-specific by Westinghouse, bounding loads were defined to fracture mechanics analyses loads; these bounding loads were used in the for flaw-induced fracture that were performed to determine anywhere within the primary the potential system main loop piping.
3.2 Fracture Mechanics Analysis An elastic-plastic fracture that large margins against mechanics analysis was performed to demonstrate stainless steel primary pipingdouble-ended pipe break.would be maintained for PWR
subjected to large Postulated that contains a large Postulated crack and determine (1) if the postulatedloadings. Key tasks in the analyses were tois the load, and (2) if any additional flaw would grow larger on stable and not result in a crack growth that mighttheoccur application of performed using axial and complete circumferential break. The would be bending associated with the facilities loads that are upper bounds analysis was identified in Table 1. For of the loads analytical purposes, TABLE 1 Operating Facilities**
Included in Westinghouse A-2 Owners Group Haddam Neck*
D. C. Cook No. 1 & 2 R. E. Ginna Point Beach No. 1 & 2 H. R. Robinson San Onofre No. 1 Surry No. 1 & 2 Turkey Point No. 3 & 4 Yankee Rowe *
Zion No. 1 & 2 Fort Calhoun
- Seismic requirements did not exist for these plants.
- The Owners Group list of ties included a foreign operating facili- No. 2 over which the NRC facility, Ringhals authority. Thus, we made has no regulatory no formal judgments regarding this facility.
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the circumference,-was
.a throughwall crack, seven inches in length around where the bounding bending postulated to exist in the pipe at the section sufficiently large so that it is moments and axial forces occur. This flaw during normal operation. (As would be very unlikely to exist undetected coolant system degradation discussed in NUREG-0691 (Ref. 8), no PWR primary has been detected to date.)
of a numerical yalue The fracture mechanics analysis required determination for the growth, or extension, of for a parameter that represents the potential system loads. This parameter a crack in a pipe that is subjected to specific as J. The J integral is is called the J integral (Ref. 1) and is denoted where the section containing the typically employed in fracture evaluations to the loading. Extension or flaw undergoes some plastic deformation due value of J reaches a critical value growth of an existing flaw occurs when the as JIC
called J initiation, which normally is denoted it is necessary to evaluate When extension of the existing crack is predicted, a stable manner or if the crack this extension and determine if it occurs in in a doubled-ended break.
will extend in an uncontrolled manner and result extension be evaluated to assess The NRC staff requires *-a predicted crack the Owners Group evaluated the stability. To comply with this requirement, stability concept and the tearing predicted crack extension using the tearing tearing stability concept is used modulus stability criterion (Ref. 2). The tearing. This mechanism can when the mechanism for flaw extension is ductile materials in the Owners Group's be expected to prevail for the primary piping following sections. The tearing facilities which are discussed further in the stability of crack extension and modulus is the parameter used to measure the as is denoted as T. Tearing modulus is defined E (1)
dJ
da ao0
to produce a specified increment where dJ indicates the increment of J needed state, of crack extension at any given load and crack E is the material elastic.modulus, and the sum a is the material flow stress defined as one half of the material yield and ultimate strengths the values.of J and T are first To determine the margin against fracture, loads and specified crack calculated for the structure using the applied analysis create the potential geometry. The values obtained from the structural Japp' and T applied or Tapp'
for fracture and are denoted as J applied, or is determined experimentally from The resistance of the structure to fracture between J and crack extension.
materials test data that show the relationship or J-R, curve. From this curve This relationship is called the J resistance, to unstable crack extension, the material tearing modulus, or the resistance J level greater than is obtained and is denoted as Tmat. At any specified JIc' stable crack extension wilt occur when
- 6- Tmat > Tapp The amount by which Tmat exceeds T is a measure of the margin against unstable crack extension or, break upon application of the in this case,.the margin against a double-ended loading to the flawed pipe.
Topical report WCAP 9558 contains
-determine J the results of the analyses app and Tapp'. The value of J app was determined fromperformed to plastic analysis using a finite an elastic- on the bounding load conditions, element computer code. The analysis was based throughwall crack, and a lower the postulated seven-inch circumferential
0
600 F. The value of Tapp was bound material stress-strain curve obtained using previously developed obtained at methods contained in Reference analytical
3.
The material J-R curves used the postulated loading and flaw to determine if crack growth would conditions and to define values occur under defined in WCAP 9558 for base of Tmat are metal carbon steel safe-end is discussed and in WCAP 9787 for weld metal. The questions (Subject Document in the Westinghouse response No. 3). A summary of the scope to ACRS
testing follows. of the materials
3.3 Materials Testing Program Base metals representative of Owners Group were selected for those in plants included in the Westinghouse Group have wrought stainless testing. All plants in the steel Westinghouse Owners has centrifugally cast stainless primary coolant piping except steel piping. one, which Westinghouse selected three heats steel for testing. Westinghouse of cast and three heats of wrought stainless strate that the tensile and also conducted tests of fracture toughness properties weld metals to demon- are comparable to those determined of the weld metal system. for the base metal in the primary piping A survey of quality assurance welds in each of the plants files was conducted to identify in the Owners Group and to define the primary piping each weld, such as the welding the details of treatment, and other pertinent process, electrode size and material, thermal matrix of representative weldinginformation. Based on the survey results, a representative welds was fabricated parameters was established -and a set of six The welds were then radiographically using typical 2 :5-inch-thick base plate.
Compact tension and tensile examined and heat treated where specimens were machined from applicable.
each weld and tested.
Tensile tests were conducted rates for five of the six heats at 600*F using conventional and material was tested at conventional of base materials. The sixth dynamic loading heat of base specimens were tested at conventionalloading rates only. Weld metal tensile loading rate tests were not loading rates for each weld.
conducted for the weld specimen. Dynamic
-7- material fracture resistance were generated J-resistance (J-R) curves to measure using compact tension specimens at conven- by multiple specimen testing at 6001F five of the six heats of base metal.
tional and dynamic loading rates for heat of base metal and the weld materials J-resistance curves for the sixth rates only. The conventional load generated at 6001F using conventional were performed in accordance with the procedures rate testing and J calculations were the dynamic toughness test, Westinghouse presented in Reference 4. To perform at predetermined displacements, thus allowing used a procedure to stop the tests from multiple-specimen dynamic teiting.
development of a J-resistance curve at conventional and dynamic loading A minimum of five specimens were tested The base metal specimens were machined rates for each of the base metal heats. that the crack would grow in the circumferen- from pipe sections and oriented so estimated 3 J c and Tmat values for tial direction of the pipe. Westinghouse each of the heats of materials tested.
from the slopes of the best-fit The values of JIc and Tmat were estimated then line through the data points for each base metal heat. Tmat was straight variation effects of crack extension using a adjusted to account for the nonlinear suggested by Ernst, et al. (Ref. 5). For of the incremental correction scheme exhibited a large amount of scatter and, the fast rate tests, the data points data points to estimate JIC or Tmat' A
in some cases, there were not enough test for each weld metal using the same minimum of three specimens were testedmetal testing. All of the weld metal procedure that was used for the base band of the base metal data points except data points fell within the scatter metal. The data points for the Inconel those for the welds with Inconel filler than any of the other base or weld metals.
weld indicated much higher toughness made no attempt at points, Westinghouse Because of the small number of data the weld metals; however, the weld metal estimating JIc or dJ/da values for lines to demonstrate trends comparable data points were fitted with straight to the base metal.
3.4 Leak Rate Calculations in Section 4.1 for defining To comply with the NRC criteria specifiedperformed to define the relationship were postulated flaw size, calculations area. The leak rate calculations were between leak rate and crack opening to to show that a postulated throughwall crack was large enough leakage performed at normal operating conditions by produce leaks that could be detected detect primary system leakage.
detection devices normally used to using the method developed by Fauske The leak rate calculations were performed this method was augmented to include (Ref. 6) for two-phase choked flow; An iterative computational scheme frictional effects of the crack surface. opening area and flow rate the sum of to the was used such that at a given crack drop was equal the frictional pressure momentum pressure drop (Ref. 6) and pressure to atmospheric (i.e.,
system the pressure drop from the primary
2250 - 14.7 psia,.
- 8- To calculate the frictional pressure estimated from fatigue-cracked stainless drop, the relative surface roughness tions were performed for a 7-inch-long steel specimens. The leak rate was
2250 psi pressure; for conservatism, circumferential throughwall crack calcula- the bending stress was assumed to at to zero for this analysis. The be equal leak rate calculated was approximately
10 gpm.
Although leak rate calculations, especially for small cracks, are uncertainties, the leak rate calculation subject to
-generated laboratory data (Ref. scheme was correlated with previously
7) and compared with service data previously detected in the PWR feedwater frem leakage tion line at Duane Arnold. In lines at D. C. Cook and the BWR
recircula- rate is sufficiently large so asspite of the uncertainties, the calculated leak normal operation. Further discussionto have a high probability of detection of the leak rate analyses is presented during the Westinghouse response to ACRS in of this evaluation. questions, the third report listed on page one
4.0 Evaluation
4.1 NRC Evaluation Criteria The evaluation of the integrity margin against ductile rupture of PWR primary system piping is based on the and resistance to fracture for a throughwall flaw and loading conditions. postulated induced fracture, the staff required To determine the potential for flaw-
(1) included an explicit crack tip the usq of analysis methods that growth of an existing crack, and parameter, (2) predicted the potential for
(3) determined if any predicted crack exten- sion would'occur in a stable manner.
fact that crack extension in ductile These requirements, coupled with ductile tearing, led the staff piping material likely will result the to use the J integral based tearing from concept as the basis for our evaluation. stability the associated tearing modulus The tearing stability concept and- previously by the staff and foundstability criterion (Ref. 2) have been. evaluated acceptable for use in the evaluation piping. of LWR
The specific criteria used with integrity of PWR primary system the tearing stability analysis to evaluate the piping and determine if adequate flaw-induced failure and pipe rupture margins against are maintained include the following:
4.1.1 Loading - The loading consists and thermal) and the loads associated of the static loads (pressure, deadweight tions. with safe shutdown earthquake (SSE)
condi-
4.1.2 Postulated Flaw Size - A
postulated to exist in the pipe large circumferential throughwall flaw is postulated throughwall flaw is wall. The circumferential length of the thickness or (2) the flaw lengthto be the larger of either (1) twice the wall that corresponds to a calculated of 10 gallons per minute (gpm) leak rate at normal operating conditions.
Although this safety evaluation system piping at the PWR facilitieshas been written exclusively for the primary LWR piping is system specific and listed in Table 1, cracking potential in concerning the generic application some additional comments are appropriate of the assumed flaw sizes used in the piping
a~nalyses. References 8 and 9 indicate that piping systems other than PWR
primary systems have some service history of observed cracking. For these 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 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 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 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 restraining 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 also can apply to this piping-to-vessel interface.
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 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 margi-ns against unstable crack extension exists, the NRC staff concludes that the condition T. > 3T
should be satisfied at the applied J level. mat - app
4.1.4 Applicability of Analytical Method The J-integral and tearing modulus computational 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)
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b > 25 J (2)
a and O
w = dJ b >> 1 w a 3(3)
where b = characteristic structural dimension, in this instance pipe wall thickness;
'IO = material flow stress;
and dJ = slope of the J-R curve at any given value of J.
da 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 w >> 1 is some- what indefinite. Generally, a range of w 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 values of b and w) also may be sufficient will have to be demonstrated to be so by additional data. These data are but now available for the piping materials considered in this investigation. not
4.1.5 Net Section Plasticity The ASME Code specifies margins for pipe stress relative to material yield ultimate strengths at faulted loading conditions. Because very large flaws and may significantly reduce the net load carrying section of the piping, analyses 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 that there are adequate margins against net section plastic failure. ensure
4.2 Evaluation Results
4.2.1 Loads The loads used to perform the fracture mechanics analyses for the primary include: piping axial tension: 1800 KIPS (includes 2250 psi pressure load), and bending moment: 45,600 in-KIPS.
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.
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loads indicated in Table 1, the enveloping With the exception of several plants pressure, and safe shutdown earthquake include those from deadweight, thermal, (pressure, deadweight, and thermal)
were (SSE) conditions. The static loads absolutely with the SSE loads.
combined algebraically and then summed axial loads and bending moments that The exceptions noted in Table 1 reportedloads (i.e., thermal, deadweight, and are comprised of only normal operating loads associated with the SSE, the major internal pressure) and did not include Our evaluation is predicated on inclusion contributor to the bending moment. Yankee and Yankee Rowe are being of the SSE loadings. However, Connecticut Evaluation Program (SEP) and are committed evaluated as part of the Systematic RCPB, safe shutdown systems, and engineered to perform se-ismic analyses of their spectra that will be available in the near safety features using site-specific is scheduled for 1983. Confirmation of future. The completion of such analyses await the extension under SSE loading will the margins against unstable crack loop piping for these two facilities.
seismic analysis of the RCPB main loads, including the analytical models, The development of the envelopingwere reviewed and approved by the staff assumptions, and computer codes, each Owners Group plant and were not reviewed during the licensing process for find that these loads, therefore, are upper again as part of this effort. We application in the fracture mechanics bound loads and are acceptable forpiping.
evaluation of the RCPB main loop
4.2.2 Materials Properties loading conducted at conventional and fast Tensile Tests - Tensile tests wereconventional loading rates for the weld metals.
rates for the base metals and at and unambiguous. A comparison of These tests are relatively straightforward and fast loading rate tests indicated the results from the conventional and decreased percentage in elongation increased yield and ultimate strengthsthe weld with the Inconel filler metal, for at faster loading rates. Except for the weld materials were comparable the yield and ultimate tensile strengths yield Inconel weld demonstrated a comparable to those for the base metal. The the base metals. With the exception of the but higher ultimate strength than reported for the weld materials were Inconel weld, the percent elongationsthe, base materials, indicating lower significantly less than those for relative ductility for the weldments.
test base metals in the plants and thewere The tensile properties for the actual indicating that the test materials program materials were compatable, Similarly, the Westinghouse survey representative of the in-plant materials.
was comprehensive and the weld specimens of weld materials and techniques representative of welds in the plants.
fabricated for testing should be national standard Fracture Toughness Testing - Currently, neither an NRC nor a various methods
3 curves, therefore exists for establishing J c or J-resistance the by different laboratories. All fracture toughness testing in are employed specimen using the multiple compact tension Westinghouse program was performed
4.
procedure outlined in Reference
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This procedure is the basis for the proposed JIc test procedure considered by ASTM Committee currently being determining JIC' The proposed E-24 and is generally considered acceptable test procedure recommends for mining J-Integral values and calculations for deter- several criteria for ensuring tion. These criteria include valid JIC determina- considerations of specimen size and data evaluation.
J-Integral Formulation - The for the compact tension specimensexpression used by Westinghouse has been shown to overestimate for calculating J
J because the experimental the value of crack growth and plasticity.data are not corrected for the nonlinear effects of calculated values of Tmat' The effect of this overestimate In order to account for these is to increase applied a correction scheme effects, Westinghouse has reviewed this scheme and based on work by Ernst, et al. (Ref. 5). The NRC
found it to be acceptable.
Specimen Size and Geometry
- Equations 2 and 3 in Section limitations to the applicability 4.1.4 specify certain analysis techniques. Because of the J-Integral and tearing of the high toughness of the instability all of the tests satisfied heats sampled, not curve, discussed later in both of these criteria. However, a lower bound J-R
this evaluation. This lower bound section, was developed for the purpose of this equations 2 and 3 over most curve typically meets the of requirements of higher levels of J where the the range of analysis. The exception is for by equation 2. However, the specimen dimensions were not adequate as specified the base metals and 2.5 inchesspecimen thickness of 1.65 inches to 2 inches for thickness of the primary coolantfor the weld metals approximate the actual ness simulates the restraint piping (2.5 inches). This similarity the piping toughness can condition in the neighborhood of a crack in thick- be represented by the materials so that test data.
Side grooving of specimens is a related subject of interest.
increases the degree of triaxiality Side shown to result in straighter in the crack tip stress field grooving and has been are desirable when J-resistancecrack fronts during crack extension. Side grooves unloading compliance test curves are developed using or when the single specimen heavy section structures such the data are applied in the evaluation of dimensions used in these tests as pressure vessels. However, since the specimen conclude that the J-resistance approximate the full thickness of the pipes, we grooves are acceptable. curves developed from specimens without side Dynamic Tests - The proposed for quasi-static testing rates. testing procedure used by Westinghouse is intended in the Westinghouse program Dynamic toughness tests that were conducted full understanding of dynamichave not previously been performed. Although a currently is not available, fracture toughness in the the elastic-plastic regime that the materials consistently significant result of the dynamic tests was initiation (higher JIC) at demonstrated greater resistance faster loading rates. However, to crack two heats of wrought stainless it is noted that the faster loading rates. steel exhibited lower estimated T. values mat at
- 13 -
proposed Based on our review of the materials test data, we conclude that the is accept- J-resistance curve test procedure referenced in the subject documents able for determining JIC and Tmat' Although the tests conducted did not specimens strictly conform to the criteria recommended in Reference 4, the test of the and procedures are judged to realistically represent the performance T values actual piping systems. In general, the reported ranges of JIc and are acceptable as representative of the structures and materials under consideration.
behavior, To perform a generic analysis and account for variations in material bound J-R
the staff used the data supplied by the Owners Group to define lower bound curves curves.for the piping materials. The data indicated that two lower of the were warranted. One lower bound curve was constructed by a composite for the wrought and weld data while the second lower bound curve was defined analyses cast material. These two lower bound curves were then used with the crack described in the next section to evaluate the margin against unstable extension for wrought and cast stainless steel piping.
4.2.3 Fracture Mechanics Evaluation were We have reviewed the elastic-plastic fracture mechanics analyses that submitted by the Owners Group. Our review included independent calculations that were performed to evaluate the acceptability of the Owners Group's conclusions.
unstable To demonstrate that the postulated throughwall flaw would not sustain first crack extension during the postulated loading, finite element calculations applied.
were performed by the Owners Group to determine Japp as a function of of 1800
bending moment with a constant axial force equal to the bounding value and bending moment provided a convenient kips. The relationship between J
plants means to associate the potential for crack extension with the individual listed in Table 1.
between We have performed independent calculations to verify the relationship Japp applied bending moment. Our calculations are approximate and are based and on elastic methods corrected for plasticity associated with the loading are the presence of the postulated flaw. While our confirmatory calculations are approximations, they do demonstrate that the Owners Group calculations prevail accurate at lower loads where elastic or small-scaleyielding conditions Further, and are conservative at larger loads where plastic deformation occurs.
analysis the Owners Group elastic-plastic analysis is conservative because theapplied was performed essentially for a section of pipe as a free body with end loads equal to the bounding loads. This is the limiting (conservative) be in a condition relative to system compliance; a pipe in a real system wouldcrack less compliant situation and would have lower potential for unstable extension.
- 14 -
Based on the J app :31ues calculated for the Owners Group by Westinghouse the lower and 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 sufficient postulated loads to cause extension of the postulatedin7 Table 1,have circumferential throughwall flaw. The loads at the remaining facilities-inch-long not high enough to produce extension of the postulated flaw. are Of the seven facilities where crack extension was predicted, one has cast stainless steel piping. Because of the differences intoughness properties between the wrought, weld, and cast materials, it was and tensile necessary to construct two distinct J-R curves. One curve was constructed from while the second was constructed from a composite of the weld and cast material wrought data.
To determine ifthe crack extension predicted for the seven facilities be stable, the Owners Group was required to determine the applied would tearing modulus, Tapp. The value of Tapp was calculated using the methods described inReference 3.
We have performed independent calculations to verify the Owners Group Tapp calcula- tions using the same methods employed inour Japp computations.
Again, our results indicate that the Owners Group calculations are conservative. Based calculated values of Tapp and the values of Tmat obtained from on the the J-R curve, we find that large margins against unstable crack extension exist facilities with predicted crack extension for the postulated flaw for the seven bending loads. sizes and We also have reviewed the method of analyses that have been performed the leak rate from the postulated flaw size for normal operating to estimate These calculations were performed to satisfy a staff requirement conditions.
that leak detection capability be included, at least qualitatively, inthe Based on our review of the leak rate calculations, we conclude piping analyses.
that lations presented by the Owners Group represent the state-of-the-art the calcu- be used to qualitatively establish the leak rate for compliance and can with current staff criteria. The leak rate has been determined to be approximately at normal operating conditions and represents, within reasonable 10 gpm limits of accuracy, detectable leakage rates at operating facilities with their available leakage detection systems or devices. For the purposes of this evaluation, isno need to backfit Reaulatory Guide 1.45 to require seismic qualification there such leakage occurs during normal operating conditions. since Based on our review, we have determined that all the facilities with the exception of the two facilities without seismic analyses, listed in Table 1.
satisfy the acceptance criteria defined.in Section 4.1. Compliance with criteria in Section 4.1 ensures that a large margin .against unstable the acceptance crack extension exists and that the potential for pipe break in the main ficiently low to preclude using it as a design basis for defining loops is suf- loads at the facilities listed in Table 1. In addition, the facilities structural do not have seismic analyses are found to be conditionally acceptable that the seismic analyses are completed and the loads are defined. until Our conditional acceptance is based on: (1) our estimate that the seismic loads to be higher than those listed for the other facilities in Table are not likely
1, (2) the wide margin against unstable fracture that exists at the maximum by Westinghouse, and (3) the low probility that large loadings will moments reported to completing the seismic analyses. occur prior
- 15 -
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 piping material; a somewhat lower limit would apply for cast material because 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 w for the test specimens and the primary piping is within the specified 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 w 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 show that lower values of w 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 exceed 42,000 in-kips.
As indicated in Section 4.1, the staff's evaluation criteria are designed to ensure that adequate margins exist against both unstable flat extension and net section plasticity of the uncracked pipe section. Both conditions are evaluated because either may be associated with pipe failure depending on the specific pipe load, material, flaw, and system constraint conditions.
Because there may be significant variations or uncertainties associated with these variables, the staff criteria do nQt 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 assess accurately by analysis because the tough materials used in LWR primary
- 16 -
piping typically produce data that fail to satisfy the size restrictions equations (2) and (3) at the very high J levels where failure would of expected to occur. be 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 against pipe failure. The staff also has reviewed the Owners Group's margin plastic analysis and data to provide additional information relative elastic- against failure. Based on this review, we conclude that, for the to margin conditions evaluated in this application, the limiting condition is associated section plasticity rather than unstable crack extension and that the with net against net section plastic failure is approximately 2.3 relative to margin the
42,000 in-kip limit and the postulated 7.5-inch circumferential throughwall flaw. This margin also can be translated into an estimate of margin size of about 5, i.e., the throughwall flaw size corresponding to on flaw plastic failure at 42,000 in-kips would be about 38 inches long or net section
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 piping 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 established by the staff to ensure that the potential for 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 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 facilities.
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.
- For all the facilities listed in Table 1, the actual moment is less than
42,000 in-kips and the Japp is less than Jmat for each facility.
- 17 -
3. The criteria used to ensure that adequate margins against breaks includes the potential to tolerate large throughwall flaws without unstable-crack extension so that leakage detection 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 detection systems should be sufficient to provide adequate margin to detect the leakage from the postulated circumferential throughwall flaw 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 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.
- 18 -
4. The additional information provided by Westinghouse in response to ACRS
questions does not alter our 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 U.S. Nuclear Regulatory Commission Reportthe Subject of Tearing Instability,"
NUREG-0311, August 19777
3. Tada, H., et al., "Stability Analysis Piping Systems," U.S. Nuclear Regulatory of Circumferential Cracks in Reactor June 1979. Commission Report NUREG/CR-0838,
4. Clarke, G. A., et al., "A Procedure for Fracture Toughness Values Using J Integral the Determination of Ductile and Evaluation, JETVA, Vol. 7, No. 1, Techniques," Journal of Testing January 1979.
5. Ernst, H. A., et al., "Estimations on J Integral and Tearing Modulus T
from Single Specimen Test Record," presented on Fracture Mechanics, Philadelphia, at the 13th Material Symposium PA, June 1980.
6. Fauske, H. K., "Critical Two-Phase, Steam Heat Transfer and Fluid Mechanics Institute, Water Flows," Proceeding of the University Press, 1961. Stanford, California, Stanford
7. Agostinelli, A. and Salemann, V., "Prediction Five Annular Clearances," Trans. ASME, of Flashing Water Flow Through July 1958, pp. 1138-1142.
8. U. S. Nuclear Regulatory Commission, "Investigation Incidents in Piping in Pressurized Water and Evaluation of Cracking September 1980. Reactors," USNRC Report NUREG-0691,
9. U. S. Nuclear Regulatory Commission, Stress-Corrosion Cracking in Piping of"Investigation and Evaluation of Report NUREG-0531, February 1979. Light Water Reactor Plants," USNRC
10. Begley, J. A. and Landes, J. D., in Fracture American Society for Testing and Materials, Analysis, ASTM STP 560,
1974, pp. 170-186.
11. Hutchinson, J. W. and Paris, P. C., "Stability Crack Growth," Elastic-Plastic Fracture, Analysis of J-Controlled for Testing and Materials, 1979, pp. ASTM S.TP 668, American Society
37-64.
four Class 1£ cables/wires are cable/wires. PG.L stated that the following environmentally qualified:
installed outside containment and have been Cable/Wire Qualification Document
1. Raychem Flametrol Test Report EM-1030; September 24, 1974
2. Okonite EPR/Hypalon Okonite. Letter Report; October 14, 1974
3. Okonite XLPE Engineering Report 367-A; January 7, 1983
4. Rockbestos XLPE Test Report S.D. 24408-5; March 3,10983 been installed outside containment which No other types of Class 1E cables haveenergy line breaks. These four types of potentially can be subjected to0 high 480 Vac between lines for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
cables have been tested to 540 F with staff reviewed the first two qualification All four types passed the test. The Flametrol cable had been qualified as reports and concluded that the Raychem cable had been demonstrated to be stated; however. the Okor.ite P?9/HvDalon subsequent discussions with the licensee, qualiied for only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Based on the staff at the PG&E offices in Sen including an audit of documentation by the statf determined:
Francisco on December 19 and 20, 1983 and therefore, are not subject to
1. The cables are enclosed in conduit direct jet impingement;
on those conduits that are essential
2. The consequences of jet impingement the staff under the same effort targets are currently being reviewed by 4.3.5;
discussed under open item 29 in Section
0 is based on the maximum temperature
3 The qualification temperature of 540 Fpostulated break; and of the steam in the pipe prior to the at a temperature of 540'F. The operator
4. The cables are qualified for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />swithin less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
will identify and isolate the break by letter prior to Mode 2 (criticality).
The licensee will submit the above information the staff review and evaluation of the infor- Based on this commitment and based on that this followup item is mation during the audit, the staff concludes resolved.
Followuo Item 15: Protection for CRVPS
that PG&E will revise the FSAR to The staff stated in SSER 18 (page C.4-17) line break analyses on the CRVPS. In incorporate results of moderate energy the following basis and schedule Board Notification 83-179 the staff provided for closeout of this item:
breaks indicated that PG&E
"The IDVP review of moderate energy line by not i.cluding the had failed to meet its licensing commitment line break analysis. PG&E
CRVPS in the criginal moderate energy that only one CRVOWS elec- provided a subsequent analysis indicating break icen.ifted by the trical train is affected by the postulated in the reaundant electrical IDVP. Wher combined with a single failure C.4-8 Diablo Canyon SSER 20
Enclosure 2 Regulatory Analysis of Mechanistic Fracture Evaluation of Reactor Coolant Piping A-2 Westinghouse Owner Group Plants
1. Statement of the Problem
2. Objective
3. Alternative
4. Cbnsequences 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. Major Assumptions IV. Conclusions B. Impact on Other Requirements C. Constraints
5. Decision Rationale
6. Implementation Attachment: Leak Before Break Value-Impact Analysis
Regulatory Analysis of Mechanistic Fracture Evaluation of Reactor Coolant Piping A-2 Westinghouse Owner Group Plants
1. Statement of the Problem results The problem of asymmetric blowdown loads on PWR primary systems (DEGB) at from postulated rapid-opening, double-ended guillotine breaks include specific locations of reactor coolant piping. These locations annulus the reactor pressure vessel (RPV) nozzle-pipe interface in the selected (reactor cavity) between the RPV and the shield wall plus other ruptures break locations external to the reactor cavity. These postulated to the could cause pressure imbalance loads both internal and external core primary system which could damage primary system equipment supports, melt cooling equipment or core internals and thus contribute to core frequency.
1975, was This generic PWR issue, initially identified to the staff in in detail designated Unresolved Safety Issue (USI) A-2 and is described in NUREG-0609 which provides a pressure load analysis method acceptable to the staff.
Owner The plants to which this analysis applies are the A-2 Westinghouse Group plants identified in Enclosure 2.
2. Objective The objective of this proposed action is to demonstrate that deterministic in fracture mechanics analysis which meets the criteria evaluated DEGB,
Enclosure 2 is an acceptable alternative to (a)postulating a modifications (b) analyzing the structural loads, and (c) installing plant
-2- to mitigate the consequences in order to resolve issue A-2. Demonstrating by acceptable fracture.mechanics analysis that there is a large against unstable extension of a crack margin in such piping, (leak before break)
contingent upon satisfying the staff's.leak detection criteria, establish a technical justification will for the identified plants to be exempted from-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 the blowdown asymmetric pressure reducing
'loads and the need to modify equipment supports to withstand those loads as determined in plant specific reported in WCAP-9628 and WCAP-9748, analysis
"Westinghouse Owners Group Asymmetric LOCA Loads Evaluation" (Evaluation of DEGB outside and inside the reactor cavity respectively).
4. Consecuences 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
- 3 -
to this enclosure.
plants has been completed by PNL and is attached suggested in the February The V-I assessment uses methods and data Assessment (PNL4646)
1983 draft of.proposed Handbook for Value-Impact Power Plant Safety and in NUREG/CR-2800, "Guidelines for Nuclear The nominal estimate Issue Prioritization Information Development."
and conclusions of the results,-major assumptions, uncertainties, III, and IV below. The assessment are discussed in Sections II,
are included in the table results of the upper and lower estimates in Section IV below.
Exposure II. Values-Public Risk and Occupational A. Results installing The estimated reduction in public risk for equipment supports additional pipe restraints and modifying pressure as necessary to mitigate or withstand asymmetric
3' man-rem total for blowdown loads is very small, only about Similarly, the the nominal case for all 16 plants considered.
with accident reduction in occupational exposure associated estimated to total avoidance due to modifying the plants is result from the less than 1 man-rem. These small changes of 1x1O-7 estimated small reduction in core-melt frequency modifying the plants.
events/reactor-year that would result from for installing However, the occupational exposure estimated increase by and maintaining the plant modifications would in occupational
11,000 man-rem. Consequently, the savings far exceed exposure by not requiring the plant modifications risk and avoided the potentially small increase in public the accident exposure associated with requiring modifications.
B. Major Assumotions and accident The above estimated changes in public risk by examining avoided occupational exposure were obtained melt from WASH-1400 accident sequences leading to core
- 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 DEGS frequencies for large primary system piping used in the analysis. These frequency estimates range from an upper estimate of 10-5 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 paper on nuclear and non-nuclear pipe reliability data in iAEA-SM-218/11, dated October
1977 by S. H. Bush which indicates a rance of
10-4 to 10-6 per reactor-year.
Additional data in the paper indicates that 10is may be 100
times too high for the pipe size being considered in Isle A-2.
An intermediate or nominal estimate of 4xO-' per reactor- year for primary system piping outside the reactor cavity and 9x10 S/reactor-year for piping inside the reactor cavity
- 5 -
are based on Report SAI-O01-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 v 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
12 over a plant is estimated to have a probability of 7x10
lifetime. For this analysis the lower estimate 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 Ix10 4 /reactor-year. However, the upper estimate of
5 /reactor-year is consistent with WASH-1400 median V0-
-
assessment pipe section rupture data. A review of the
16 plants under consideration indicates there are an
"Later work (to be published) by LLL indicates that an indirect seismically induced DEGB (e.g., earthquake-induced failure of a polar crane or heavy from component support-steam generator or RC pump) is more probable ranging to 10-10 /rea:tor-year with a median of 10- /reactor-year for plants east
7 L:-;
paper c the Rockies. Since the nominal DEGE frequency obtained from the IAEA
a::roximates the median indirect DEGB frequency, the direct DEGB estimate of
7x:0-'2 over a plant lifetime was used for the lowewr estimate.
- 6- N-_
average of 10.3 sections of primary system piping per reactor. Multiplying this value by 1
8.8x10- 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 Only Class I primary system Miscellaneous primary and of pipe pipe with nuclear grade QA
secondary system piping and ISI
of various classifications y-ye of failure Double-ended guillotine (DEG) Circumferential and long- break only itudinal breaks, large cracks Failure location Selected primary system break Random system break locations locations 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
- 7 -
of radioactive isotopes as used in WASH-1400, the meteorology at a typical Midwestern site (Byron-Braidwood), a uniform population density of 340 people per square-mile (which is an average of all U.S. nuclear power plant sites) and no evacuation of population. They are based on a 50-mile release radius-model.
5. The change in occupational exposure associated with accident avoidance assumes 20,000 man-rem/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 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 (WCAP 9628). The modifications consisted of added RPV
nozzle-pipe restraints and substantial modification of all steam generator and pump supports. The occupational exposures for these 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 RPV nozzle-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 assumed to be required at 14 of the 16 plants and would incur about 200 man-rem total.
- 8 -
III. hmDacts - Industry/NRC Costs
- Property Damaqe A. Results The estimated industry costs to install plant modifications to withstand asymmetric pressure-loads is about $50 million.
It is, also estimated that power replacement costs would be 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 work 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 mcdifiaclt;
k -nd power replacement cost indicated above.
It is estimated that it would cost NRC about $BOOK in staff review effort if plant modifications to withstand asymmetric pressure loads were to be installed.
If they are not installed and this cost is saved, then it is estimated that NRC cost would be $400K 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.
It is estimated that installing plant modifications to withstand asymmetric pressure loads would avoid public prooerty damage costs due to an accident by S24K to S36"
- 9 -
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 $15K 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
1. The costs for installing the plant modifications were determined by separating the plants into two groups.
The cost for the first group of three plants which require extensive modifications used an estimate submitted by San Onofre Unit 1 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 Loads," and from industry estimates including informal estimates from DC Cook. The estimates were adjusted to 1982 dollars.
2. The cost estimates 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 risk upper and lower bound variations are related to Indian Point 2 and Palo Verde values calculated from NUREG/CR 2723.
- 10 -
3. Power replacements costs were based on an assumed $300K
per plant outage day.
IV. Conclusions The results of the Value-Impact assessment are summarized in the table below. In the 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 plant 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-rem) and industry and NRC
costs (SilO million - including
$60 million power replacement cost). Three additional observations are worth noting:
a) the uncertainty bounds show net positive benefits for 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 z.his assessment.
The detailed PNL Value-Impact assessment is attached to this enclosure.
LEAK BEFORE BREAK VALUE-IMPACT SUMMARY - TOTAL FOR 16 PLANTS
Dose (man-rem) Cost (S)
Nominal Lower Upper Nominal Lower Upper Factors . Estimate Estimate Estimate Estimate Estimate Estimate values (man-rem)
-3.4 0 -37 - - -
Public Health occupational Exposure -0.8 0 -30 - -
(Accidental)
Occupational Exposure +1.1xlO' +3500 +3.2X10 4 - -
_10perational)
Values Subtotal +1.1x10 4 +3500 +3.2x104 - -
ImDacts (S)
Industry iml men - - - -50x106 -25x106 -75x106 station Cost a
- -6.5x10 5 -3.3x105 -9.8x10 5 Industry Operating Cost - -
NRC Development and Implementation-Cost(b) - - - -4.Ox105 -2.0x105 -6.0x10 5 Power Replacement Cost - - - -60x106 -30x106 -90x106 Public Property - +2.4x104 0 +2.6x106
- - +1.5x104 0 +4.6x105 Onsite Property -
impact Subtotal - - - -110x10 6 -55x10 6 -165x106 (a) Does not include industry costs expended to date to prepare plant asymmmetric pressure load analyses and pipe fracture mechanics analysis.
(b) Does not include NRC cost expended to date to develop issue (NUREG-0609) and to evaluate Westinghouse pipe fracture mechanics analysis.
- 12 -
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 Owner 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 piping DEG break.
6. Implementation The steps and schedule for implementation of the proposed action are discussed in Sections 3.5 thru 3.9 and
5.2.1, 5.2.2, 5.2.3 of
.ncIcsure S.
LEAK BEFORE BREAK VALUE-IMPACT ANALYSIS
,I. 1iNTPOlUCTIOrI
This report presents a value-impact assessment of the consequences of exempting Westinghouse A-2 Owners Group plants from having to Install modifi- cations to mitigate asymmetric blowdown loads in the primary system.. This.
assessment uses methods suggested in the Handbook for Value-Impact Assessment (Heeberlin et al..1083) and data developed for safety issue prioritization (Andrews et al. 183). 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).
The proposed action will efficiently allocate public resources in the generation of electric power and avoid occupational dose with only small increments to public risk. Modification of plant designs to accommodate 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. Previouslyprimary unanalyzed loads on primary system components have the potential to alter system configurations or damage core cooling equipment and contribute to core melt accidents. For postulated pipe breaks in the cold leg, asymmetric pressure changes could take place in the annulus between the core barrel and vessel the RPY.
Decompression could take place on the side of the reactor pressure (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- pression of the vessel. For breaks in RPV nozzles, the annulus between the reactor and biological shield wall could become asymmetrically pressurized, resulting in additional horizontal and vertical external loads on the reactor vessel. In addition, the reactor vessel is loaded simultaneously by the effects of strain-eneroy 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 because of the more-rapid decompression of the upper plenum. Similar asymmetric forces could and also be generated by postulated pipe breaks located at the steam generator reactor- coolant pump. The blowdown asymmetric pressure loads have been analyzed and reported in WCAP-9628 (Campbell et al. 1080) and WCAP-974R (Campbell et al.
1079), "Westinghouse Owners Group Asymmetric LnCA Loads Evaluation."
2.n PRnPnSED ACTIO?! AenD PnTEN!T'AL ALTERNA'TIVES
It is proposed that Westinghouse A-2 Owner Grnup plants listed in blow- Erclosu-e 2 be exempted from plant modifi.yations to mitigate asyrnetric I
down loads-to pr ary system components. This ation of public risk, occupational dose and proposal is based on consider- -
cost impacts. The alternative would be to require each operating PWR to add system component supports to withstand the blowdownpiping restraints and primary asymmnetric pressure loads.
Public risk reductions for installing/modifying asymmetric blowdown loads are small. Extensive equipment to mitigate properties and crack propagation by industry analyses of pipe material (WrAP-9558 and WCAP-9787, Campbell et al, 1982 and 1981) and the NRC indicate that through-the-wall cracks are extremely unlikely. catastrophic failures without plants upgrade leik detection systems, as necessary, It is proposed that these detection capabilities. This will to provide adequate leak allow before they propagate to major failures. cracks Plant to be identified and repaired modifications occupational dose and inspection time for primary-system would increase reduction in the frequency of core-melt accidents components. The accident doses as a result of the plant modifications and avoidance of post- is not significant.
Cost impacts for equipment to mitigate asymmetric dependent. In the worst case, they cost many millions blowdown loads are plant replacement power purchases and are of questionable of dollars, require considered can handle asymmetric loads with few feasibility. Some plants will realize cost savings for the proposed action. changes. However, all plants
3.0 AFFECTED DECTSION FACTORS
Causes Causes Quantified Unquantified2() No Decison Factors Change Chanae Chance Public Health X
Occupational Exposure (Accidental) X
Occupational Exposure (Routine) X
Public Property X
nnsite Property X
Regulatory Efficiency Improvements in Knowledqe X
Industry 'molementation Cost X
Industry Operation Cost X
NRC Development Cost X
NIrC Implementation Cost X
I'DC Operation Cost X
X
'a Tn tbis context, "unquantified' means not readily estimated in dollars.
2
AENT SUMMARY - Total for 16 P ts VALUE-IMPACT LSSE.
Nominal Lower Upper Estimate Estimate Estimate Decision Factors Values(a) (man-rem)
-3.4 0 -37 Public Health n -30
Occupational Exposure -(n.8 (Accidental) 3.2Ej'A
1.lE+A 3500
Occupational Exposure (Operational)
Regulatory Efficiency N/A
Improvements in Knowledge N/A
1.1Et4 3500 3.2E+4 Total Quantified Value IQpacts(b)
Industr lmplementation -1.6ES
Cost) -1.1E+8 -5.3E+7
-6.5E+5 -3.3E+S -9.BE+5 Industry Operating C?8j 0
NRC Development CostJ Q .
-4.QE+5 -2.OE-5 -6.0E+5 NRC Implementation Cost 0 6 NRC Operation Cost 0
2.4E+4 h 2.6E-6 Public Property &.FE-S
1.5E+4 0
Onsite Property
-1.1E+8 -5.3E+7 -1.6E+8 Total Quantified Impact NRC goals. Principle (a) A decision term is a value if it supports of safety.
among these goals is the regulation as a result of the (b) ImDacts are defined as the costs incurred indicate cost savincs.
proposed action. Negative impacts date (fracture to (c) Does not include industry cost expended load analyses).
mechanics and plant asymmetric pressure Replacement power costs of S6AM are included. asymetric loads (Hosford (dW Does not include NRC costs to evaluate(Campbell 1982).
198M) or industry fracture mechanics N'!A = Not Affected
.n UOLIANTI!FIE- RESIDUAL ASSESSMENT
in the assessment of this action.
There are no uncuantified decision factors A.0 DEVELODt',E!T (IFOIALIF!f.T0IN
A. Public Health the effects of exernotirn
^ risk a.nalvsis wis performed to assess rioii'ications to nit4:ate V'rest5n-, use G-,.:~A-2 owner nroup -lants from
3
asyrrnetric blowdcwn\<.ds on primary by examining WAS- *InO accident sequencessystem component,' This was accomplished rupture and large LOCAs. leading to core melt from vessel u l For this analysis, it was assumed large LOCA can occur either inside that a double-ended guillotine (DEG)
to the "standard" stresses caused or outside the reactor cavity. In by a large LOCA (depressurization addition coolant inventory), the DEG break and loss of can have additional effects:
1. If the DEG break occurs inside the asymmetric blowdown which displaces reactor cavity, it can cause an other pipes or the vessel itself. the reactor vessel, possibly rupturing
2. If the DEG break occurs anywhere in the primary loop, it can cause asymmetric blowdown which 1) displaces an becomes uncoolable and/or 2) fails the core such that its geometry (ECCS) piping through dynamic blowdown needed emergency core cooling system forces.
Three sources of data were used to develop estimates of DEG break bilities used in this analysis. proba- These probability estimates range from upper estimate of IE-S breaks per an
7E-12 breaks in a reactor lifetime.reactor year down to a lower estimate of The upper estimate is based on reliability data (Bush 1977). This a study of nuclear and non-nuclear pipe failures per reactor year. Failures data indicates a range-of 1E-4 to 1E-6 ruptures, disruptive and potentially considered include leaks, cracks, to 1E-6 are representative of disruptive disruptive. Bush indicates values of this analysis as an upper estimate. failures. A value of 1E-5 was used1E-5 Additional data presented by Bush in cates that this value may be 100 indi- times too high for the pipe sizes considered in the proposed action. being An intermediate or nominal estimate Fullwood 1976) that modeled crack is based on a study by SAI (Harris propagation and fatigue stresses. While the study in piping that is subject to the aporoach and data are not plant was done for Combustion Engineering plants, service inspection as a metnod to specific. Conservatively ignoring discover and repair cracks prior in- propagation, SAI reports DEG break to unstable primary system and 9E-8/py in the frequency estimates of 4E-7/py for the life for a two loop plant (Figure reactor cavity averaged over a 40-year plant
23, Harris and Fullwood 1976).
The lower estimate of a Lnr.A was atories (Lu et al. 1981) usinc simulation developed by Lawrence Livermore Labor- crack propagation in primary system techniques to model direct effects other cyclic stresses. Indirect piping due to thermal, pressure,seismic on effects such as external mechanical and were not included. Results indicate damage likely than breaks and that breaks leaks are several orders of magnitude have more lifetime. This value is essentially a probability of 7E-12 over a plant additional lower estimate calculationszero for risk calculation purposes, so no were performed.
4
oIt is acknowlzI1ednat both the upper and nominal estimate DEG break frequencies used this analysis are less than the WASH-1400 large LOCA median frequency of lE-4/reactor-yr. However, the upper estimate of IE-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 average of
10.3-sections of primary system piping/reactor. Multiplying this value by
8.8E-7 rupture/section-year for large (>'") pipe obtained from Table II .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-1OO that support use.of a lower pipe break frequency. These factors are tabulated below:
Westinghouse A-2 Factor Owners Group Plants WASH-14n0 Large LOCA
Pipe size >30 inches diameter - >6 inches diameter Pipe material - austenitic stainless steel - carbon steel and stainless steel System and class - only class I primary system - miscellaneous primary and of pipe pipe with nuclear grade QA 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 primary system - random system *break break locations locations Leak detection - LOS capability to detect - no requirement or provision system (LnS) leak in A timely manner for leak detection to maintain large margin Against unstable crack extension Jt 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-1 00 are assumed. These sequences are as follows (Table V.3-14, dominant only):
RC-aL (PLR-1) with frequency = 2E-12/py RC-Y- (PWR-2) with frequency = 3E-11 Ipy RC-6 (PWR-2) wi th frequency = lE-il/py RC-6 (PWR-2) wish frequency = IE-12/py R-a (PWI?- ' with frequency = IE-9/py pot (PWR-7) with frequency = 1E-7!py WASM-l1n0O assumes a vessel rupture frequency of 'E-7 py. Replacing this with
9-8I/py t"he nominal estmeate frequency or in-cavity asynmetric blowdown auto- D
?tieal causinc melt in a way analogous to<_.ssel rupture) resu~lts in- the san- oreviouS equence frequencies.
^cse es:imates for the release catecories were and -ssu.i-nc the quantities of radioactive isotopes derived using the CRAC code and Guidelines used in WASH-
14On, he -;-teorology at a typical midwestern site (Byron-Braidwood), a uniform popu!Ation density of %O people per square-mile U.S. nuclear power plant sites) and no evacuation (which is an average o' all of population. They are based or a 50-mile release radius model.
Tne nominal es.i2mate risk from the in-cavity DEG large LOCA in a two loop plant becomes:
Pisk = (2E-12/py)(f.;'C+6 man-rem) + (4 E-11/py)(4.8E+6 man-rem) +
(1E-9/py!(5r.tE.5 man-rem) T ( IE-7/py)(2300 man-ren)
= .Qo man-rer./py was assumed that asyrnetric blowdown from a DEG large LOCA outside reactor cavity does not automatically lead the systi-. failures would be needed to result in to a core-melt. Subsequent safety core-melt, although the potential for the IEG larce LOrA to cause such failures such that its geometry becomes uncoolable) directly (or displace the core still exists.
Presumably, failure of safety systems independent accounTed for in the plant design. Since the of asymmetric loading are WASHr-inn large LOCA sequence, it was assumed DEG brPak is only part of the that no risk is added by the break itself. Only safety system failures induced by unanticipated asymretric loads on equipmert or core geometry disruptions contribute to this issue.
'o calculate the contribution to core melt from breaks outside the reactor caviyV, a two-step analysis was followed.
First, the contribution to core melt Iro, -EG breks cutside the reactor cavity was additieonal fract'on of this contribution, hased calculated. Second, an analtses, ;.as cazculated to represent. the risk on previous systems interaction blowavan contribution diue to asymemeric Onlv -his fract.ion would be incurred for the PEG Dreaks were previously considered in the pro5osed action since plant design.
To estimate the risk contribution cavity, accident sequences analogous tofrom DEG breaks outside the reactor those for a large LOCA in WASH-14on assu-ed applicable. These sequences are as are follows (Table V.3-14, dominant AB-,a -IP- 11with frequency = 1E-lI/py M{;.
-1 .cr--
Vt2]- !@! @ - 5EtI/
= ir_10/PY
A-Y !?S'4-2E = I-lO'Dy
-
=2E_1. 12py
'.^_ ~~p "- ~ *~-' ~" ~ n ':_8
~2E-R/pi ~ r2,C
AF- 6 (PWR-3N " = 1E-8/py AG-o (PWR.3 " 9E-9/py
- ACO-E (PWR-40 " " 1E-11/py AD- S(PWR-5) " "E-9/py AM- L (PWR-5_) = 3E-9/py AB-C (PWR-6) " " = l-9/py AHF-E (PWR-6) " " = lE-lO/py ADF- E (PWR-6) 2E-10/py AD- C (PWR-7) " al = 2E-6/py AH- c (PWR-7) " U = IE-6/py TOTAL 3E-6/py WASH-1400 assumes a median large LOCA frequency of IE-A/py. Replacing this with 4.DE-7/py (the nominal estimate frequency of outside-of-cavity DEG large 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 dependent failures):
Risk = (1E-12/py)(5.4E+6 man-rem) + (6E-13/py)(4.RE+6 man-rem) +
(2EH-1/py)(5.&E+6 man-rem) + (4E-34/py)(2.7E46 man-rem) +
(2E-11/py)(l.OE-6 man-rem) + (5E-12/py)(l.5E+5 man-rem) +
(1.2E-8/py)(2300 man-rem)
= IE-3 man-rem/py As assessed in the report for safety issue II.C.3 (Systems Interaction) in Supp. 1 to NUREG/CR-2800 (Andrews et al. 1983), systems interactions typically contribute 10% to total core-melt frequency (and risk), with a range of 1l-
201. The types of safety system failures which could be induced directly by adverse forces from a DEG large LOCA causing asymmetric blowdown are typical systems interactions The Westinghouse G7A'P -2wors croup has provided analyses for ex-cavity breaks that indicate disru-:4c- o core geormetry is unlikely to occur (Campbell
1980) for 13 out of 16 plarts. However, to account for this possibility and that of asymmetric-blowdown-induced damaoe to safety equipment, the uoper end of the range for systems interaction contribution (20%) is assumeo applicable to estimate the risk frorm dependent failures resulting from outside-of-cavity asymmetric blowdown. Thus, the incremental best estimate risk from the outside- of-cavity DEG large LOCA with asymmetric loadings becomes:
Risk £ (n.2)(1E-3 mean-rem/py)
= 2E-4 man-rem/py Combining the two scenarios for DEG large LOCAs within and outside of the reactor cavity yields the following total risk for two loop plants:
Risk = 0.006 + 2E-4 = 0.006 man-rem/py Nominal estimate results for plants that use a two-loop corficuration were adJusted to account for the added number-of loops in some plants. A review of
7
the GTAP A-2 owne-s sup list indicates that these
3.1 loops. The r.-.inal estimate becomes 0.009 p',-.its have an average of man-rem/py.i Upper estimate risk calculations were made using those of the nominal estimates. The pipe rupture procedures similar to cated 8(% to the primary loop and 20% to the frequency of IE-5 was allo- reactor cavity by assuming the ratio of results from the SAI study. No corrections loops are necessary because this frequency is for the number of plant failure rate of 2E-6 is 20 times higher than per -plant year. The in-cavity WASH-1d00 for vessel rupture. The upper estimate cavity risk becomes:
Risk a (dE-i1Jpy)(5.&E+6 man-rem) +
(8.2E-10/py)(4.8E+6 man-rem) +
(2.0 E-8/py)(5.4E+6 man-rem) +
(2.OE-6/py)(23(O man-rem)
= 0.12 man-rem/py The upper estimate of primary loop breaks of
8E-6 is 12 times lower than WASH-1400 for large LOCAs. The upper estimate loop risk becomes:
Risk - 0.2 E(2E-11/py)(5.AE+6 man-rem) + (1.3E-11/py)(4.SE+S
(3.9E-9/py)(5.tE+6 man-rem) + (8E-13/py)(2.7E+5 man-rem) +
(S.6E-10/py)(IEji6 man-rem) + (i.EDE-l0/py)(1.5E+5 man-rem) +
(2.4E-7/pyl(2300 man-rem) man-rem) +
= 0.on man-rem/py Combining 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 remaining plant years (16 plants x 23.6 yr = in these cases by the number of
377 py) results in the industry total public risk increase due to leak before break.
Total Added Risk (man-rem)
Nominal Estimate 3.d Upper Estimate 37 Lower Estimate n A nominal estimate for the total increase proposed action was determined by summing the in core melt frequency for the
'hp reactor cavity and out-nf-cavity loop breakcontributions for breaks inside ar.Justinc for the average systems interactions and then number of looDS.
8
Core nelt inc-ase -3.1/2E9E-R + 0.2(3E-6/2501 = 1T-74/py by An upper estimate of the core-melt frequency increase was calculated and 20% of summing the contributions from reactor cavity pipe breaks (2E-06/py)
the out-cf-cavity pipe break initiated core melt accidents.
Core melt increase = 2E-6 + O.2(2E-7) = 2E-6/py Total core-melt frequency increase estimates are as follows:
Increase in Core-Melt Freauency (Events/py)
Nominal Estimate 1E-7 Upper Estimate 2E-6 Lower Estimate 0
B. Occupational Exposure - Accidental the The increased occupational exposure from accidents can be estimated as product of the change in total core-melt frequency and the occupational in core exposure likely to occur in the event of a major accident. The change in The occupational exposure melt frequency was estimated as 1E-7 events/yr. "immediate"
the event of a maior accident has two components. The first is the its short exposure to the personnel onsite during the span of the event and with the term control. The second is the longer term exposure associated cleanup and recovery from the accident.
The total avoided occupational exposure is calculated as follows:
OTO = 7NTlOA; DA= P(DIO+DLTO)
where
= Total avoided occupational dose M = Number of affected facilities
= Average remaining lifetime Pro = Avoided occupational dose per reactor-year a= Change in core-melt frequency P,) = "Imarediate" occupational dose DLTC = Long-term occupational dose.
ae conservativelv Pesults c- -he calculations ara shown below. Uncertainties nppe'r hound tii er d orooaca-ec by use of extremes (e.c.,
9
r In-. :ase in Immediate(a) Long Term(8)
Cc e Melt Occupational Total Occupational Avoi ded Frequency Dose Dose Occupational (events! Oman-ren/! (man-rem/ Exposure)
reactor-yr) event) event)
-
(man-rem)
'ominal 1E-7 1E3 Estinate 2E4 0.R
tipper Estimate 2E- 4E3 3E4 30
Lower Estimate a 0 1EA
n (a) Based on cleanup and decommissioning estimates, NUREG/CR-2601 (Murphy
1982).
C. Public Property The effect of the proposed action upon the calculated by multiplying the change in accident risk to offsite property is property damage estimate. This estimate frequency by a generic offsite results of CRAC2 calculations, assuming was derived from the mean value of
154 reactors (Strip 1982). CRAC2 includes an SST1 release (major accident), for displaced.persons, property decontamination, costs for evacuation, relocation of property through interdiction and crop and loss of use of contaminated milk losses. Litigation costs, impacts to areas receiving evacuees and institutional The damaae estimate is converted to present costs are no: included.
discount rate was also considered as a sensitivity value discounting at 10%. A 5X
case.
The following discounting formula is employed:
D= y e I _e '
I
where D = discounted value
,. = deaage estimate years before reactor begins operation;
t. = years remaining n for operating plants until end of life.
I = discount rate o. Lthis r posed action, only operating reactors are affected, anc the average rumber of years of remaining life is 23.5.
P/V = 9. The 5% discount factor equals 12.8. Therefore, the 10* discount factor tv tne number of affected facilities (l6i-to These values must be multiplied
..tion. *Upper rd lower bcunds are values for yield the total effect of m-e cnaculaed from Sz.rip (19R2). Results Indian Point 2 and 310 Vprce 3 are as follows:
10
Discounted Offsite -- Discounted Property namage Value of Additional Of'site Property [Lifetime Risk3 Offsite Property famage (S/event) (S/event) Damage (WI
- W0P - 10%W 5w Nominal 1.7E+0 1.5+10 3E+10
L2. 2.4E+4 3.BE+4 Estimate Upper Estimate 9.2E+9 8.3E.10 1.3E*11 2.6E+6 4.1E+6 Lower Estimate R.3E+8 7.5E+10 1.2E+lo n 0
D. Onsite ProDerty Thp effect of the proposed action on the risk to onsite 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. (183). 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.
D ( [ I I
fl-e-1 I (-e -(tf -t
) 53 where D = discounted value V - damage estimate m = years over which cleanup is spread - 10 years ti= years before reactor begins operation; n for operating plants t C years remaining until end of life; 0 - 2X.5 years I= discount rate c 10Q or 5%.
For this proposed action, the IlM discount factor equals 5.7 and the 5%
discount factor equals 11. To obtain the total effect of the action, the per- reactor results are multiplied by the number of affected facilities (16). The uncertainty bounds given in the table reflect a 500 spread which was estimated to se indicative of the uncertainty level. The results are summarized below:
11
I
On -e Property Discounted Discount Value of Avoided Danace Estimate nnsite Property Onsite Property (S/event! Damage (S/event) Damaae (S)
10% 5W % 5 Nominal 1.65E+9 9.&E+9 1.8ElO 1.5E+4 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.OE+9 0 n E. Occupational Exposure-Operational Operational occupational exposure due to plant modifications is avoided by the proposedinstallation and maintenance of loads during implementation and operation. exemption to asymmetric blowdown For this analysis, plants were broken into extensive modifications and the rest. two groups; those requiring A listing of each group and assumed modifications is given in the section on implementation doses for the three plants Industry Implementation Cost. Avoided were based on a San Onofre estimate of requiring extensive modifications system pipe restraints at the RPV nozzles 2600 man-rem/plant to install primary generator supports for three loops. Some and modifying pump and steam for the proposed action to upgrade leak occupational doses will be incurred detection systems. For these plants, it is estimated that U5O man-hours per plant
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 inside containment at 45 mR/hr and would be required to install such modifications. No modifications to the this group, net avoided implementation core or core barrel were assumed. For doses were calculated as follows:
Avoided installation dose a 3[2600 - (0.0025
(80) + 0.045 (450))J
= 7700 man rem Implementation doses for -he remaining thirteen follows: 80% of total direct costs plants were estimated were assumed to be attributed to labor as in radiation zones. These costs were converted cost per man year (assumed to be MR00k) to man-hours by dividing by the year. Man-rem estimates were calculated and multiplying by 18nO man-hours/man- inside containment and 2.5 mR/hr outside by assuming dose rates of 25 mR/hr of containment. The lower value for containment work was assumed due to less extensive better equipment access. Required activities modifications and presumed
'iplementation Costs. are described further in Industry
12
Total avoide occuF .Jonal doses' due to implementatsiun, operation and maintenance are Known below. Upper and lower estimates were developed using the following model (Andrews et al. 1983):
Do~se upper - 3 dose expected Dose lower 1/3 dose expected Activity nose Avoided (man-rem)
Implementation 9700
Operation, Maintenance 840
Total 1.IE.4 Upper Estimate 3.2E+4 Lower Estimate 3500
F. Industry Implementation Cost Several levels of value to industry are seen as resulting from thp proposed action. Potential desion 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 on assumed avoided plant modifications:
Plants Requiring Extensive Modifications:
Haddam Neck Yankee Rowe San Onofre 1 Plants Requiring Some Modification:
HS Robinson 2 Zion 1,2 Turkey Point 3,4 RE Ginna Surry 1,2 Point Beach 1,2 DC Cook 1,2 Ft. Calhoun.
For plants requiring extensive modifications, data developed for modifi- cation to primary system component supports and vessel nozzle restraints by San Onofre were used (Baskin 19.80). Total reported costs were divided by three to obtain a per-loop cost. Costs for contingencies were ignored. Results are as fol 1ows:
14
- Results of thl: an),..Jsis are.as follows:
Number Direst of Avoided Ccst a Plants Dose Rate.
Activity Implementation
'S/looP) (Loops) Man-Hourstb) (R/hr) Dose fman-Rem)
Ins-tall primary shield wall restraints and inspection port modifications 98000 13 (40 )(dke) 56000 0.025 1d0o Modify reactor coolant pump supports 20000 21 )(d)
I( 6000 0.025 150
Steam generator supports 120000 4 (12 )(d) 21000 0.025 520
Calibrate leak(C)
detection system N/A 11 (f) 5000 0.025 (120)
Total
2000
(a) Stevenson 1980, except for shield wall and inspection Costs for these activities are based on industry port modificat ons.
estimates for D.C.* ook.
(b) (nirect Cost)(Humber of Loops)(18no man-hr/man-yr)(O.8)/(SI.nz/man-yr!.
(c)Avoided doses are negative for these activities because they for the proposed action. are required (d} Campbell 1979 and 198n.
(e) Ft. Calhoun was credited with 3 loops due to (f) Two plants have verified adequate leak redundant cold legs.
detection capability.
Occupational dose to maintain the modifications estimate the amount, it was assumed that two additional is also avoided. To year would be spent inside containment if the modifications man-weeks per plant- due to inspection of the modifications and additional are made. This is access to primary system components. The total time required to Cain dose fcr the owners oroLD is estimated below. Plants requiring extensive modifications totaling 56 plant-years. All other plant lives total have renaming lives
320 plant-years.
ODerational dose averted = (8 ioan-hr/py)[(56
. plant-years )(0.rE.l R/nan-hr).
(320 plant-years)(0.025 R,/man-hr)!
= 840 man-rem
13
Stevenson;
materiel and labe-.s .11 other costs listed are bases )n work by supplier and The original worK aid not appear to include engineering, NSSS
An additional Into was assumed for these costs based on util'.Y support costs. 1°* for the San Onofre data. All costs were also increased by an additional escalations between 1980 and 19f82.
All modifications would not be required at all plants. Based on Owners number of Group analyses (Campbell 1979), it was assumed that the following modifications would be performed.
Owners Group Avoided Modification Number of Plants (Loops) Cost Primary Shield Wall 13 (40) S9200K
Restraint and Inspection Port Modification
7 (21) S110OK
Reactor Coolant Pump Supports Steam Generator Supports 4 (12) S3700K
Reactor Vessel Supports 0 0
Reactor Coolant Compartment 0 0
Walls Total S14OO0K
to be Shield wall restraints and inspection port modifications were assumed to work was assumed required at all plants. Pump and steam generator support vessel supports be needed at plants identified by the owners group. Reactor them as were assumed not to be needed by any plants. Stevenson discusses are only mainly a seismic restraint. Reactor coolant compartment wall anchors required for the safe shutdown earthquake (SSE) and LOCA load combinations.
Thus they were not used in this analysis.
identified Needs for replacement power to modify remaining plants were not and steam in the available data. It was assumed for plants requiring pump be needed generator support modifications that some replacement power would it. was assumed that one half of' the (four plants). For this analysis,
-ncrermentalout-ae time of San Onofre would be needed or 20 days. Total outage replacement power at S30OK/day total S2oM.
days would be 80. Costs for for plants Ccsts for modifying 7eak detection systems are assumed the same extensive modifications. It was recuiring some modification as for plants with Costs for this work assumed thAt only 11 of the 13 plants need upgrading.
.c-.a S2.RE-5.
as
'.-Wavoided ccsts for plants with some modifications were calculated f1 13vws:
16
, g - Per-Loop;_3sts (SK)_
Direct Costs (materials, field costs)
A/E Support 90J
NSSS Supplier Support 333 Utility Support 716 Escalation (1979-1982) 166
740
Total
2856 In addition, Baskin reports that 40 days of purchased. At S30nK/day (Andrews et al. 1983), replacement power would be costs are S12M per plant. the total replacement power It is conservatively assumed that all to their leak detection systems. This may three plants will require upgrading measurement systems and revisions to technical include calibration of current flow upgrades are based on labor estimates of 0.25 specifications. Costs for these yr,total costs are S25K/plant. man-yr. At SlO0K per man- Total implementation costs for the three plants were calculated as follows:
Implementation costs (Total Number of Loops)(Avoided Cost per Loop)-
(Number of Affected Plants)r(Replacement Power +
Avoided Cost) - (Leak Detection Costs)!
(11)(S2.86E+6) + 3CSI.2E+7 - S2.5E.-4)
= S6.7E+7 Implementation costs for the remaining plants (Stevenson 1980) and industry estimates including are derived from UCPL-153an indicated below: San Onofre. Results are Modification Cost Primary Shield Wall Restraint and Inspection Port Modification (Hot and Cold Leg) S23nK/loop Reactor Coolant Pump Supports S 52K/loop Steam Generator Supports S311K/loop Reactor Vessel Supports S 19K/loop Reactor Coolant Component Walls.
S230K/pl3nt The shield wall restraints and inspection ruptures in the reactor cavity. These costs port modifications are to control based on estimates for DC Cook units and are were escalated in 19S2 dollars assumed to include all overheads,
15
Avoided NRC lmplem.,ation Support Costs:
16 plants (O.25 man-yr/plant e S100,000/man yr) = S4.DE+5 tipper Estimate = S6.OE+B
Lower Estimate = S2.OE+5 No additional NRC costs during operations are expected.
7.0 CONCLUSIONS
The summary results for the value-impact assessment are shown below. 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 cost. The upper estimate is very positive. The following observations can also be made:
o This action did not address costs of core and core support modifications.
Adding these costs would increase the negative impact of the exemption.
o 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 primarily occupational dose during equipment installation. This dose is being weighed against statistical estimates of public and occupational dose for rare events.
o Cost results are not sensitive to discount rates used in this analysis.
JVul2'Y of Value-Impact Assessment Value !r,.n-rem) impact (S)
Nominal Upper Lower Est. Est. Est. Nominal Est. Uoper Est. Lower Est.
10_ 5% 10__ 5I o1t%
- .*rd 3.27-4 35nZ 1.1 + -l.lE+R -1.6iE-~E+8 -1.6E+S -5.3E.,7 is
Net Avoided Impst *ta:ion Costs ' Primary Systemic ificatiors Replacement Podrr - Leakage Detection Systems.
Sl.LE+7 + S2.4E+7 - S2.SE+5
= S3.IE+7 To gene-ate upper and lower estimates for costs, it was mAtes are within WO of the nominal estimate. Results assumed that esti- tation costs are summarized below: for industry implemen- Plants with Extensive Modifications S6.7E.7 Plants with Some Modifications S.RE+7 Total S1. 1E+8 Upper Estimate S1.6E+8 Lower Estimate $5.3E.7 G. Industry Operation and Maintenance Costs Industry avoided operation and maintenance costs were the assumption that. additional restraints will result developed based on in additional inspections and restrict access to steam generators, reactor coolant nozzles. Based on the values used for occupational dose pumps and reactor is assumed to total PO man-hours/plant-year. At S100K/man-year estimates, this labor wk/man-yr, the annual cost is S4540/plant. The present and t4 man- for 16 plants over 23.5 years with upper ard lower estimates value of this quantity are as follows:
Discount Rate
10 6, Present Value of Operation and Maintenance Costs = $6.5ES5 1.OE-6 Upper Estimate = S9.8E+5 1.5E+6 Lower Estimate = S3.3E+5 5.OE+5 H. NRC Implementation Suonort costs NRC Avoided Implementation costs are estimated to be to review plant modifications. This is partially offset 0.5 man-year of labor man-vear to review leak detection system upgrades and by an estimate of 0.25 technical specifications. Net NJC cost savings are as revisions to plant follows:
17
Murprv, E. S., and :,. M. Holter. 1982. Technology, Safety and Costs c' Decornissicnir Reference Light Water ,eactors Following Postulated EC-.e:s. FlU7,2E;R-26nl, PaciTic Nor:hvest LaDonratory, Ricrnanc,
'cc hasninctcn.
Stevenson, J. 0. 1980. Cost and Safety Margin Assessment of the Effects of Desicn 'or Combination of Large LOCA and SSE Loads. UCRL-15340, Lawrence Livermore Laooratory, Livermore, Catifornia.
Strip, D. Rt. 1982. Estimates of the Financial Consequences of Nuclear Power Reactor Accidents. MlUREGi/CR-2723, Sancia National Laboratories, Albuquerque, New Mexico.
' zo oU.s. ooyzREMN PRIUTING OFFICE a 19U4 O-421-637/139
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