ML20116J306
| ML20116J306 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 04/23/1985 |
| From: | Miraglia F Office of Nuclear Reactor Regulation |
| To: | DUKE POWER CO. |
| Shared Package | |
| ML20116J309 | List: |
| References | |
| NUDOCS 8505020562 | |
| Download: ML20116J306 (25) | |
Text
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7590-01 UNITED STATES MUCLEAR REGULATORY CO*1ISSION In the Matter of
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Docket No. 50 a14 1
(Catawba Nuclear Station, Unit 2) i EXEMPTION I.
On July 24, 1972, Duke Power Company tendered an application for licenses to construct Catawba Nuclear Station, Units 1 and 2 (Catawba or the facility) with the Atomic Energy Commission (currently the Nuclear Regulatory Commission or the Comission).
Following a public hearing before the Atomic Safety and Lic.ensing Board, the Commission issued Construction Permit Nos. CPPR-116 and CPPR-117 permitting the construction of Units 1 and 2, respectively, on August 7, 1975.
Each unit of the facility is a pressurized water reactor, con-taining a Westinghouse Electric Company nuclear steam supply system, located in York County, South Carolina.
On March 21, 1979, Duke Power Company tendered an application for operating licenses for each unit of the facility. On January 17, 1985, the Director of the Office of Nuclear Reactor Regulation issued a full power license for Catawba Unit 1.
Catawba Unit 2 remains in the licensing review process.
By letters dated May 15, 1978, and April 8, 1981, Duke Power Company trans-mitted two applications for amendment to Construction Permit No. CPPR-117 to add respectively North Carolina Municipal Power Agency No. 1 (NCMPA-1) and Piedmont Municipal. Power Agency (PMPA) as co-owners of Catawba Unit 2.
By letters dated l
October 19, 1978, and November 24, 1981, the staff has amended CPPR-117 to reflect the two changes in ownership.
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8505020562 850423 PDR ADOCK 05000414 A
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II.
The Construction Permit issued for constructing the facility provides, in pertinent part, that the facility is sub.iect to all rules, regulations and orders of the Commission. This includes General Design Criterion (GDC) 4 of Appendix A to 10 CFR 50.
GDC 4 requires that structures, systems and components important to safety shall be designed to accommodate the effects of, and to be compatible with, the environmental conditions associated with the normal operation,' main-tenance, testing and postulated accidents, including loss-of-coolant accidents.
These structures, systems and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, discharging fluids that may result from equipment failures, and from events and conditions.
outside the nuclear power unit.
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In a submittal dated December 20, 1983, the applicants enclosed Westinghouse Report MT-SME-3166 (Reference 1) containing the technical basis for their request to: (1) eliminate the need to postulate circumferential and longitudinal pipe breaks in the RCS primary loop (hot leg, cold leg, and cross-over leg piping);
(2) eliminate the need for associated pipe whip restraints in the RCS primary loop and eliminate the requirement to design for the structural effects associ-ated with RCS primary loop pipe breaks including jet impingement; and (3) elin-inate the need to consider dynamic effects and loading conditions associated with previously postulated primary loop pipe breaks.
These effects include blowdown loads, jet impingement loads, and reactor cavity and subcompartment pressuriza-tion. Attachment 3 to the December 20, 1983, submittal identified the primary loop break locations and the erection status of the associated pipe whip restraints.
By. letter dated February 14, 1985, the applicants updated the installation status for those restraints.
Furthermore, by letter (fren W. H. Owen, Duke, to H. R. Denton, f!RC) dated April 17, 1985, the applicants 6
- withdrew that portion of the December 20, 1983, exemption request that related to the leak-before-break effects on reactor cavity and subcompartment pres-surization. Alsc, by the same letter the applicants requested that a partial exemption to GDC-4 be granted for the first two cycles of operation.
The applicants also stated in their submittals that employment of the leak-before-break concept would not eliminate pipe breaks in the RCS primary locp as a design basis for the following: (1) containment design; (2) sizing of Emergency Core Cooling System; (3) environmental qualification of eouipment; and (4) supports for heavy components.
Based on its review of the applicants' December 20, 1983, submittal, the NRC staff requested additional information and provided comments on the reports
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-(References 1 and 9) which were transmitted to the applicants in the form of questions by NRC letter dated April 10, 1984 (Reference 2).
By a submittal dated May 11, 1984, the apolicants responded to the staff's questions, providing a new report identified as Westinghouse Report WCAP-10546 (Reference 3).
In the same submittal, the applicants requested an exemption from a portion of the requirements of GDC 4 of Appendix A to 10 CFR 50.
By letter dated September 14, 1984, the applicants submitted an analysis of the occupational radiation dose reduction which, together with the information contained in the December 20, 1983, and May 11, 1984, submittals, constituted the value-impact analysis for Catawba Unit 2.
The technical information con-tained in references 1 and 3, together with the value-impact analysis, provided a comprehensive justification for reouesting a partial exemption from the requirements of GDC 4.
From the deterministic fracture mechanics analysis contained in the tech-nical informa;:on furnished, the applicants contend that the postulated
4-doubleLended guillotine breaks (DEGB) of the primary loop coolant piping in Catawba Unit 2 will not occur and, therefore, need not be considered as a design basis for installing protective devices, such as pipe whip restraints and.iet impingement shields, to guard against the dynamic effects associated with such postulated breaks.
No other changes in design requirements are addressed within the scope of the referenced reports; e.g., no changes to the definitier of a LOCA nor its relationship to the regulations addressing design requirements for ECCS (10 CFR 50.46), containment (GDC 16, 50), other engineered safety features and the conditions for ervironmental qualification of equipment (10 CFR 50.49).
III.
The Commission's regulations require that app.licants provide protective measures "...against dynamic effects, including the effects of missiles, pipe whipping, and discharged fluids, that may result from equioment failures and from events and conditions cutside the nuclear power unit." (GDC 4) Protective measures include physical isolation from postulated pipe rupture locations if feasible or the installation of pipe whip restraints, jet impingement shields or compartments.
In 1975, concerns arose as to the asymmetric loads on pres-surized water reactor (PWR) vessels and their internals which could result from these large postulated breaks at discrete locations in the main primary coolant loop piping. This led to the establishment of Unresolved Safety issue (USI)
A-2, " Asymmetric Blowdown loads on PWR Primary Systems."
The NRC staff, after several review meetings with the Advisory Committee on Reactor Safeguards (ACRS) and a meeting with the NRC Committee to Review Generic Requirements (CRGR), concluded that an exemption from the regulations would be acceptable as an alternative for resolution of USI A-2 for sixteen facilities owned by eleven licensees in the Westinghouse Owners' Group (one of
. these facilities, Fort Calhoun has a Combustion Engineering nuclear steam supply system). This NRC staff position was stated in Generic Letter 84-04, published on February 1, 1984 (Reference 4).
The generic letter states that the affected licensees must justify an exemption to GDC 4 on a plant-specific basis. Other PWR applicants or licensees may request similar exemptions from the requirements of GDC 4 provided that they submit an acceotable technical
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basis for eliminating the need to postulate pipe breaks.
The acceptance of an exemption was made possible by the development of advanced fracture mechanics technology. These advanced fracture mechanics techniques deal with relatively small flaws in piping components (either postulated or real) and examine their behavior under various pipe loads. The objective is to demonstrate by deterministic.analy'ses that the detection of small flaws by either inservice inspection or leakage monitoring systems is assured long before the flaws can grow to critical or unstable sizes which could. lead to large break areas such as the DEGB or its equivalent. The concept underlying such analyses is referred to as " leak-before-break" (LBB).
There is no' implication that piping failures cannot occur, but rather that improved knowledge of the failure modes of piping systems and the application of appropriate remedial measures, if indicated, can reduce the probability of catastrophic failure to insignificant values.
Advanced fracture mechanics technology was applied in topical reports (References 5, 6 and 7) submitted to the staff by Westinghouse on behalf of the licensees belonging to the USI A-2 Owners Group. Although the topical reports were intended to resolve the issue of asymmetric blowdown loads that resulted from a limited number of discrete break locations, the technology m_m_
- advanced in these topical reports demonstrated that the probability of breaks occurring in the primary coolant system main loop piping is sufficiently low such that these breaks need not be considered as a design basis for requiring installation of pipe whip restraints or jet impingement shields. The staff's Topical Report Evaluation is included as Enclosure 1 to Reference 4.
Probabilistic fracture mechanics studies conducted by the Lawrence Liver-more National Laboratories (LLNL) on both Westinghouse and Combustion Engineer-ing nuclear steam supply system main loop ofping (Reference 8) confirm that both the probability of leakage (e.g., undetected flaw growth through the pipe wall by fatigue) and the probability of a DEGB are very low. The results given in Reference 8 are that the best-estimate leak probabilities for Westinghouse nuclear steam supply system main loop piping range" from 1.2 x 10-8 to 1.5 x 10-7 per plant year and the best-estimate DEGB probabilities range from 1 x 10-12 to 7 x 10-12 per plant year.
Similarly, the best-estimate leak probabilities for Combustion Engineering nuclear steam supply system main loop piping range from 1 x 10-0 per plant year to 3 x 10-8 per plant year, and the best-estimate DEGB probabilities range from 5 x 10-14 to 5 x 10-13 per plant year. These results do not affect core melt probabilities in any significant way.
During the past few years it has also become apparent that the requirement for installation of large, massive pipe whip restraints and jet impingement shields is not necessarily the most cost effective way to achieve the desired level of safety, as indicated in Enclosure 2, Regulatory Analysis, to Reference 4.
Even for new plants, these devices tend to restrict access for future inservice inspection of piping; or if they are removed and reinstalled for inspection, there is a potential risk of damaging the piping and other safety-related components in this process.
If installed in operating plants, high occupational i
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n radiation exposure (GRE) would be incurred while public risk reduction would be very low.
Removal and reinstallation for inservice inspection also entail significant ORE over the life of a plant.
IV.
The primary coclant system of Catawba, Unit 2, described in Reference 3, has four main locps each comprising a 33.9 inch diameter hot leg, a 36.2. inch diameter crossover leg and 32.14 inch diameter cold leg piping. The material in the primary loop piping is cast stainless steel (SA 351 CF8A).
In its review of Reference 3, the staff evaluated the Westinghouse analyses with regard to:
the location of maximum stresses in the piping, associated with the combined loads from normal operation and the SSE; potential cracking mechanisms; size of through-wall cracks that would leak a detectable amount under normal loads and pressure; stability of a " leakage-size crack" under normal plus SSE loads and the expected margin in terms of load; margin based on crack size; and the fracture toughness properties of thermally-aged cast stainless steel piping and weld material.
The NRC staff's criteria for evaluation of the above parameters are delineated in its Topical Report Evaluation, Enclosure I to Reference 4, Section j
4.1, "NRC Evaluation Criteria", and are as follows:
(1) The 1oading conditions should include the static forces and moments (pressure, deadweight and thermal expansion) due to normal operation, and the forces and moments associated with the safe shutdown earth-quake (SSE).
These forces and moments should be located where the highest stresses, coincident with the poorest material properties, and induced for base materials, weldments and safe-ends.
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- (2) For the piping run/ systems under evaluation, all pertinent information which demonstrates that degradation or failure of the piping resulting from stress corrosion cracking, fatigue or water hammer is not likely, should be provided.
Relevant operating history should be cited, which includes system operational procedures; system or component modifica-tion; water chemistry parameters, limits and controls; resistance of material to various forms of stress corrosinn, and performance under cyclic loadings.
(3) A through-wall crack should be postulated at the highest stressed locations determined from (1) above.
The size of the crack should be large enough so that the leakage is assured of detection with adequate margin using the minimum instal' led leak detection capability when the pipe is subjected to normal operational loads.
(4)
It should be demonstrated that the postulated leakage crack is stable under normal plus SSE loads for long periods of time; that is, crack growth, if any, is minimal during an earthquake. The margin, in terms of applied loads, should be determined by a crack stability analysis, i.e., that the leakage-size crack will not experience unstable crack growth even if larger loads (larger than design loads) are applied.
This analysis should demonstrate that crack growth is stable and the final crack size is limited, such that a double-ended pipe break will not occur.
(5) The crack size margin should be determined by comparine the leakage-size crack to critical-size cracks.
Under normal plus SSE loads, it should be demonstrated that there is adequate margin between the leakage-size e
9-crack and the critical-size crack to account for the uncertainties inherent in the analyses, and leakage detection capability. A limit-load analysis may suffice for this purpose, however, an elastic-plastic fracture mechanics (tearing instability) analysis is preferable.
(6) The materials data provided should include types of materials and materials specifications used for base metal, weldments and safe-ends, the materials properties including the J-R curve used in the analyses, and long-term effects such as thermal aging and other limitations to valid data (e.g. J maximum, maximum crack growth).
V.
Based on its evaluation of the analysis contained in Westinghouse Report WCAP-10546 (Reference 3), the staff finds that the applicants have presented an acceptable technical justification, addressing the above criteria, for not installing protective devices to deal with the dynamic effects of large pipe ruptures in the main loop primary coolant system piping of Catawba, Unit 2.
This finding is predicated on the fact that each of the parameters evaluated for Catawba Unit 2 is enveloped by the generic analysis performed by Westinghouse in 9eference 5, and accepted by the staff in Enclosure 1 to Reference 4.
Specifically:
(1) Although the moment associated with the highest stressed location in the main loop primary system piping (which for Catawba Unit 2 occurs in the cross over leg piping) is lower than the bounding moment used by Westinghouse in Reference (5) for the hot leg oiping, it is slightly e
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higher than that established by the staff as a limit (i.e., a moment of 42,000 in-kips in Enclosure 1 to Reference 4). However, this is compensated for in that the pipe diameter and thickness are larger than those' analyzed by Westinghouse and the staff for the reference case. Thus, ttie resultant net stresses are within the bounds established by the staff in Enclosure 1 to Reference (4). The Catawba loads are 1,864 kips (axial) and 43,407 in-kips (bending moment).
(2) For Westinghouse plants, there is no history of cracking failure in reactor primary coolant system loop piping. The Westinghouse reactor coolant system primary loop has an operating history which demonstrates its inherent stability. This includes a. low susceptibility to cracking failure from the effects of corrosion (e.g., intergranular stress cor-rosion cracking), water hammer, or fatigue (low and high cycle). This operating history totals over 400 reactor-years, including five (5) plants each having 15 years of operation and 15 other plants with over 10 years of operation.
(3) The results of the leak rate calculations performed for Catawba, using an initial through-wall crack of 7.5 inches are identical to those of to Reference (4). The Catawba plant has an RCS pressure boundary leak detection system which is consistent with the guidelines of Regulatory Guide 1.45, and it can detect leakage of one (1) gpm within one hour.
The calculated leak rate through the postulated flaw is at least 10 gpm. Therefore, the Catawba plant leak detection system is capable of detecting leaks one-tenth that of the calculated leak rate.
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(4) The margin in terms of load for Catawba Unit 2, based on fracture mechanics analyses for the leakage-size crack under normal plus SSE loads, is within the bounds calculated by the staff in Section 4.2.3 of Enclosure 1 to Reference 4.
Based on a limit-load analysis, the load margin is about 2.4 and based on the J limit discussed in (6) below, the margin is at least 1.3.
(5) The margin between the leakage-size crack and the critical-size crack was calculated by a limit-load analysis. Acain, the results demon-strated that a margin of at least 3 on crack size exists and is within the bounds of Section 4.2.3 of Enclosure 1 to Reference (4).
(6) As an integral part of its review, the staff's evaluation of the material properties data of Reference (9') is enclosed as Appendix I to this Safety Evaluation Report.
In Reference 9, data for ten (10) plants, including the Catawba Units, are presented, and lower bound or " worst case" materials properties were identified and used in the analysis performed in the Reference (3) report by Westinghouse. The applied J for Catawba in Reference (3) was less than 3000 in-lb/ine and, hence, the staff's upper bound on the applied J (refer to Appendix I, page 6) was not exceeded.
In view of the analytical results presented in Reference 3 and the staff's evaluation findings related above, the staff concludes that the probability or likelihood of large pipe breaks occurring in the primary coolant system loop of Catawba Unit 2 is sufficiently low such that protective devices associated with postulated pipe breaks at the eioht (8) locations per loop in Catawba Unit 2 o
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orimary coolant system-(as specified in the applicants' submittals of December 20, 1983, and May 11, 1984) need not be installed.
Pcwever, in order to provide the Commission with an opportunity to consider the long tern aspects of the NRC staff's recent acceptance criteria of the " leak-before-break" approach, this -
exemption is limited to a period extending until the completion of the second refueling outage cf Catawba Ur.it 2, pending the outcome of Commissinr rulemaking on this issue.
The applicants' request does not affect the design bases for the contain-ment, the emergency core cooling system, the environmental qualification of equipment for Catawba Unit 2, or the supports for heavy equipment, and does not propose to alter the design basis of reactor cavity and subcompartment pres-suriration from that originally performed which wa's based on a limited displace-ment DEGB.
The staff agrees that this schedular exemption does not affect these matters.
The staff also reviewed the value-impact analysis provided by the appli-cants in their December 20, 1983, May 11, and September 14, 1984, submittals for not providing protective devices against postulated reactor coolant system loop pipe breaks to assure as low as reasonably achievable (ALARA) exposure to plant personnel.
Consideration was given to design features #or reducing doses to personnel who must operate, service and maintain the Catawba Unit 2 instru-mentation, controls, equipment, etc.
The Catawba Unit 2 value-impact analysis shows that the elimination of protective devices for RCS pipe breaks will save an occupational dose for plant personnel of approximately 600 person-rem for Catawba Unit 2 over its operating lifetime.
The staff review of the analysis shows it to be c reasonable estimate of dose savings.
Therefore, vith respect m
. to occupational exposure, the staff finds that there is a radiological benefit to be gained by eliminating the need for the protective structures.
VI.
In view of the staff's evaluation findings, conclusions, and recommenda-tions above, the Commission has determined that, oursuant to 10 CFR 50.12(a),
the following exemotion is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest.
The Commission hereby approves the limited schedular exemption from GDC 4 of Appendix A to 10 CFR Part 50, to permit the applicants not to install protective devices and not to consider dynamic effects and loading conditions as detailed in Part II of this exemption associated with postulated pipe breaks of the eight (8) locations'per loop in the Catawba Unit 2 primary coolant system, for a period ending at the completion of the second refueling outage, pending the outcome of rulemaking on this subject.
Pursuant to 10 CFR 51.32, the Commiss'on has determined that the issuance of the Exemption will have no significant impact on the environment (50 FR 15802 ).
Dated at Bethesda, Maryland, this 23'd day of April 1985.
This Exemption will become effective upon date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION h2L'Ah Vs Frank J. Mi
- glia, ty Director Division of Licensing Office of Nuclear Reactor Regulation l
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REFERENCES fli Vestinghouse Report MT-SME-3166, " Technical Bases #or Eliminatinc large Primary Loco Pipe Ruptures as the Structural Design Basis for Catawba, Units 1 and 2," November 1983, Westinghouse Class 9 proprietary.
(21 Letter to H. B. Tucker of Duke Power Coroany, "Recuest for Additinnal Information Concerning Leak-Before-Break Analysis for Catawba Nuclear Station Unit 2, dated Aoril 10, 1984 f31 Westingbouse Report UCAP-10546, " Technical Aases #or Eliminating Larce Prinary looo Pipe Ruptura as the Structural Oesign Rasis for Catawba Unit 2," April 1984, Westinahouse Class 2 proprietary.
(4) NRC Generic Letter 84-04, " Safety Evaluation of Westinghouse Topical Reoorts Dealing with Elinination o# Postulated Rreaks in DVR Primary Main loops," February 1, 1084.
(5) Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containina a Postulated Circum #erential Throughwall Crack, VCAP-9558, Rev. ?, Mav 1981, Westinchouse Class ? proprietary.
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(6) Tensile and Touchness Properties of Primary Piping Veld Metal for Use in Mechanistic Fracture Evaluation, WCAP-9787, May 1981, Westinghouse Class ? prnnrietary.
(7) yestinghouse Response to Questions and Comments Raised by Members o' ACDS Subcomnittee on Metal Components During the Westinchouse Presentation on September 25, 1981, Letter Report NS-EPR-?519, E. P. Rahe to Darrall G.
Eisenhut, November 10, 1981, Westinghouse Class ? proprietary.
(8) Lawrence Livermore National Laboratory Report, UCRL-86249, " Failure Prob-ability o# PWR Reactor Coolant loop Pipina," by T. Lo, H. u, yoo, g, 3, Holman and C. K. Chou, Presented at the ASME PVP Con #erence and Exhibition,
. lune 17-?l, 198a, San Ontonio, Texas.
(9) Vestinghouse Report WCAP-10456, "The Effects o' Thermal Aging on the Structural Integrity of Cast Stainless Steal Piping for Westinghouse Nuclear Steam Supply Systems," November 1083, Westinghouse Class ?
oroprietary.
Notes: See next page e
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REFERENCES j
t!OTE: P!cn-proprietary versions of References 1, 3, 5, 6, 7 and 9 are available' t
in.the NRC Public Document Room as follows:
fli MT-SME-3179, non-proprietary (3) WCAP 10547 (5) WCAP 9570.
(6) WCAP 9788 (7) Non-oroorietary version attached to the lattar D.eport (91 WCAP 10457-4.
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3 APPENDIX I Evaluation of Westinghouse Report WCAP 10456, "The Effects of Thermal Aging on the~ Structural Integrity of Cast Stainless Steel Piping for Westinghouse Nuclear Steam Supply Systems" INTRODUCTION The primary coolant piping in some Westinghouse Nuclear Steam Supply Systems (NSSS) contain cast stainless steel base metal and weld metal.
The base metal and weld metal are fabricated to produce a duplex structure of delta (6) ferrite in an austenitic matrix.
The duplex structure pro-duces a material that has a higher yield strength, improved weldability and greater resistance to intergranular stress corrosion cracking than a single phase austenitic material.
However, as early as 1965 (Ref.1),
it was recognized that.long time thermal aging at primary loop water temperatures (550 F-650 F) could significantly affect the Charpy impact toughness of the duplex structured alloys.
Since the Charpy impact test is a measure of a material's resistance to fracture, a loss in Charpy impact toughness could result in reduced structural stability in the piping system.
The purpose of Report WCAP 10456 is to evaluate whether cast stainless steel base metal and weld metal containing postulated cracks will be sensitive to unstable fracture during the 40 year life of a nuclear power plant.
In order to. determine whether a piping system will behave 4
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. -Se in such a fashion, the pipe materials' mechanical properties, design criteria and method o'f predicting failure must be established.
In this evaluation, we wil1 assess the mechanical properties of thermally aged cast stainless, steel pipe materials, which are reported in Report WCAP 10456.
DISCUSSION 1.
Weld Metal Report WCAP 10456 refers to test results reported in a paper by Slama, et.al. (Ref. 2) to conclude that the weld metal in primary loop piping would not be overly sensitive to aging and that the aged cast pipe base metal material would be structurally limiting.
In the Slama report eight (8) welds were evaluated.
The tensile properties were only slightly affected by aging.
The Charpy U notch impact energy in the most highly sensitive weld decreased from 7daJ/cm2 (40 ft-lbs) to near 4daJ/cm2 (24 ft-lbs) after aging for 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at 400 C (752 F).
This change was not considered significant.
The relatively small effect of aging on the weld, as compared to cast pipe material was reported to be caused by a difference in microstructure and lower levels of ferrite in the weld than in the cast pipe material.
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3-2.
Cast Stainless Steel Pipe Base Metal Report WCAP 10456 contains mechanical property test results from a number of heats of aged cast stainless steel material and a, metallurgical study, which was performed by Westinghouse, to support a statistically based model for predicting the effect of thermal aging on the Charpy impact test properties of cast stain-less steel.
As a result of these tests and the proposed model, Westinghouse concludes that the fracture toughness test results from one heat of material tested represents end-of-life conditions for the ten (10) plants surveyed.
The ten (10) plants surveyed are identified as Plants A through J.
a.
Mechanical Property Test Results Reported in WCAP 10456 M' chanical property test results on aged and unaged cast stainless e
steel materials which were reported in a paper by Landerman and Bamford (Ref. 3), Bamford, Landerman and Diaz (Ref. 4), Slama et.'al.
(Ref. 2) were discussed in Report 10456.
In addition, Westinghouse performed confirmatory Charpy V notch and J-integral tests on aged cast stainless steel material, which was tested and evaluated by Slama et. al.
e
. The results of these tests indicate that:
(1) The fatigue crack growth rate of aged or unaged material in air and pressurized water reactor environments were equivalent.
(2) Tensile properties were essentially unaffected except for a slight increase in tensile strength and a decrease in ductili ty.
(3) J-integral test _results indicate that the J and tearing 1C modulus, T, are affected by aging.
b.
Mechanism Study in WCAP 10456 The tests and literature survey conducted by Westinghouse indicate that the proposed mechanism of aging occurs in the range of operating temperatures for pressurized water reactors and the data from accelerated aging studies can be used to predict the behavior at operating temperatures.
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Cast Stainless Steel Pipe Test s
m ThematerialsdatadiscussedinJ.heprevioussectionoftbis
- 2. evaluation were obtained from small specimens. As a consequence, s
the JR results are limited to relatively short crack extensions.
To investigate the behavior of cast stainless steel in actual piping geometry, Westinghouse performed two exper,iments, one of which was Efth thermally aged cast stainles'y, steel and the other test was identical except that the steel Yas not thermally aged.-
Each pipe tested contained a throughwall circumferential crack r
to the extent specified in WCAP 10456.
The pipe sections were closed at the ends, pressurized to nominal PWR operating l
pressure and then bending loads were applied.
t The results of the tests were very similar, in that both pipes displayed extensive ductility, and stable crack extension.
There was no observed unstable crack extension or fast fracture.
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. The results^of the Westinghouse pipe experiments indicate that cast stainicss steel, both aged and unaged, can withstand crack extensions well beyond the range of the J-R results with small specimens.
However, if crack extension'is predicted in an actual application of thermally aged cast stainless steel in a piping system, we believe that it is prudent to limit 2
the applied J to 3000 in-lbs/in or less unless further studies and/or experiments demonstrate that higher values are tolerable.
Loss of initial toughness due to-thermal aging of cast stainless steels at normal nuclear facility operating temperatures occurs slowly over the course of many years; therefore, continuing study of the aging phenomenon ray lead to a relaxation of this position.
Conversely, in the unlikely event that the total loss of toughness ar.d the rate of toughness loss are greater than those projected in this evaluation, the staff will take appropriate action to limit the values to that which can be justified by experimental data.
Because the aging is a slow process, the staff believes there would be sufficient time for the staff to recognize the problem and to rectify the situation.
However, the staff believes this situation is highly unlikely because the staff has accepted only the lower bounds of data that were gathered among ten plants encompassing the range of materials in use.
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d.
Effects of Thermal Aging on Westinghouse Supplied Centrifugally Cast Reactor' Coolant Pioing Reported in WCAP 10456
+.
The reactor coolant cast stainless steel piping materials in the plants identified in WCAP 10456 as A through J, were produced to the specification SA-351, Class CF8A as outlined in ASME Code Section II, Part A and also to Westinghouse Equipment Specification G-678864, as revised.
For these materials, Westinghouse has calculated the\\ predicted end-of-life Charpy U-notch properties, based on their proposed model.
The two (2) standard deviation end-of-life lower. limit v'alue for all the plants surveyed was s
greater than the Charpy U, notch properties of the aged reference t
materials, which Westinghouse indicates represents end-of-life
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properties for all the plants!
As a result, Westinghouse con-cluded that the amount of embrittlement in the aged reference
'J material exceed the amount projected at end-of-life for all cast 1,
stair,less steel pipe materials in Plants A through J.
1 Conclusions Based on our review of the information and data contained in Westinghouse
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' Report WCAP 10456, we conclude that:
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8-1.
Weld metal that-is used in cast stainless steel piping system is initially less fracture resistant than the cast stainless steel base metal.
However, the weld metal is less susceptible to thermal aging than the cast _ stainless steel base metal.
Hence, at end-of-life the cast stainless steel base metal is anticipated to be the least fracture resistant material.
2.
The Westinghouse proposed model may be used to predict the relative amount of embrittlement on a heat of cast stainless steel material.
The two standard deviation lower confidence limit for this model will provide a useful engineering estimate of the predicted end-of-life Charpy impact properties for cast stainless steel base metal.
3.
Since there is considerable scatter in J-integral test data for the heats of material tested, lower bound values for J and T 1c should be used as engineering estimates for the fracture resistance of the aged reference material. We believe these values should also provide a lower bound for the fracture resistance of aged and unaged weld metal.
If crack extension is predicted in an actual application of cast stainless steel in a piping system, we conclude that the r
2 applied J should be limited to 3000 in-lbs/in or less unless further studies and tests demonstrate that higher values are tolerable.
The Westingaouse pipe tests demonstrate that this may be possible.
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4.
Since the predicted end-of-life Charpy impact values for the materials in Plants A through J are greater than the value measured for.the aged reference material, the lower bound fracture properties for aged reference material may be used to determine the fracture resistance for the cast stainless steel material in Plants A through J.
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' 4' REFERENCES (1)
F. H. Beck, E. A. Schoefer, J. W. Flowers, M. E. Fontana, "New Cast High Strength Alloy Grades by Structural Control," ASTM STP 369 (1965)
(2)
G. Slama, P. Petrequin, S. H. Masson, T. R. Mager, "Effect of Aging on Mechanical Properties of Austenitic Stainless Steel Casting and Welds,"
presented at SMIRT 7 Post Conference Seminar 6 - Assuring Structural Integrity of Steel Reactor Pressure Boundary Components, August 29/30, 1983, Monterey, Ca.
(3)
E. I. Landerman and W. H. Bamford, " Fracture Toughness and Fatigue Characteristics-of_ Centrifuga11y Cast Type 316 Stainless Steel After Simulated Thermal Service Conditions.
Presented at the Winter Annual Meeting of the ASME, San Francisco, Ca.,-December 1978 (MPC-8 ASME)
(4)
W. H. Bamford, E. I. Landerman and E. Diaz, " Thermal Aging of Cast Stainless Steel and Its Impact on Piping Integrity." Presented at ASME Pressure Vessel and Piping Conference, Portland, Oregon, June 1983.
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