Safety Evaluation Supporting Conditional Approval of Fracture Mechanics Analysis of Reactor Vessel Flaw Indications in Hot Leg Nozzle to Shell Weld for UtilML20155D623 |
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05/27/1988 |
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Office of Nuclear Reactor Regulation |
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ML20155D618 |
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NUDOCS 8806150281 |
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20210D8951999-07-23023 July 1999 Safety Evaluation Accepting First 10-yr Interval Inservice Insp Plan Requests for Relief ISI-018 - ISI-020 ML20195J8951999-06-17017 June 1999 Safety Evaluation Granting Relief for Listed ISI Parts for Current Interval,Per 10CFR50.55a(g)(5)(iii) ML20207E8631999-06-0303 June 1999 Safety Evaluation Accepting Licensee 990114 Submittal of one-time Request for Relief from ASME B&PV Code IST Requirements for Pressure Safety Valves at Plant,Unit 3 ML20205A6331999-03-25025 March 1999 SER Accepting Request to Use Mechanical Nozzle Seal Assemblies as an Alternative Repair Method,Per 10CFR50.55a(a)(3)(i) for Reactor Coolant Sys Applications at Plant,Unit 3 ML20203H8591999-02-17017 February 1999 Safety Evaluation Accepting Licensee Second Ten Year ISI Program & Associated Relief Requests for Plant,Unit 3 ML20199H6261999-01-21021 January 1999 Safety Evaluation Accepting Classification of Instrument Air Tubing & Components for Safety Related Valve Top Works.Staff Recommends That EOI Revise Licensing Basis to Permit Incorporation of Change ML20196F4911998-12-0101 December 1998 SER Accepting Request for Relief ISI2-09 for Waterford Steam Electric Station,Unit 3 & Arkansas Nuclear One,Unit 2 ML20195C4841998-11-0606 November 1998 SER Accepting QA Program Change to Consolidate Four Existing QA Programs for Arkansas Nuclear One,Grand Gulf Nuclear Station,River Bend Station & Waterford 3 Steam Electric Station Into Single QA Program ML20237C5661998-08-17017 August 1998 Safety Evaluation Accepting Licensee Request for Exemption from Section Iii.O of Appendix R to 10CFR50 ML20236S9031998-07-22022 July 1998 SER Accepting Rev 19 to Quality Assurance Program for Waterford Steam Electric Station,Unit 3 ML20217P8281998-04-0707 April 1998 Safety Evaluation Accepting Relief Authorization for Alternative to Requirements of ASME Section Xi,Subarticle IWA-5250 Bolting Exam for Plants,Per 10CFR50.55a(a)(3)(i) ML20203E9371997-12-0909 December 1997 Safety Evaluation Granting Relief Request,Per 10CFR50.55a(g) (I) ML20199H3711997-11-19019 November 1997 SER Accepting Approving Request Relief from Requirements of Section XI, Rule for Inservice Insp of NPP Components, of ASME for Current or New 10-year Inservice Insp Interval IAW 50.55(a)(3)(i) of 10CFR50 ML20216E9921997-09-0404 September 1997 Safety Evaluation Accepting 970623 Request for Relief Re Authorization for Use of ASME Code Case N-416-1 & N-532,ISI Program for Listed Plants ML20141H8411997-07-30030 July 1997 Safety Evaluation Accepting Use of Code Case N-508-1 for All Four Plants for Rotation of Serviced Snubbers & Pressure Relief Valves for Purpose of Testing in Lieu of ASME Code ML20141F3881997-06-30030 June 1997 Safety Evaluation Authorizing Licensee Request for Extension of First ISI Interval to 970924 ML20148H1271997-06-0505 June 1997 Supplemental Safety Evaluation Re Conformance to Reg Guide 1.97,Entergy Operations,Inc,Waterford Steam Electric Station Unit 3 ML20149M4221996-12-12012 December 1996 Safety Evaluation Supporting Update Insvc Insp Programs to 1992 & Portions of 1993 ASME Boiler & Pressure Vessel Code, Sect XI for Licenses DPR-51,NPF-6,NPF-38,NPF-29 & NPF-47. Technical Ltr Rept Encl ML20107F5611996-04-17017 April 1996 Safety Evaluation Providing Guidance on Submitting plant- Specific Info W/Respect to IST Program Alternatives Request ML20058A3661993-11-17017 November 1993 Safety Evaluation Allowing Accumulator Level & Pressure Monitoring Instrumentation to Be Relaxed from Category 2 to Category 3 & Allowing Commercial Grade Instruments to Be Used,In Ref to GL 82-33 & Reg Guide 1.97 ML20059J1721993-11-0808 November 1993 Safety Evaluation Accepting First 10-yr Interval ISI Program Through Rev 5,except Where Relief Denied ML20127D1361993-01-11011 January 1993 Safety Evaluation Re IST Program Request for Relief.Util Proposal Complies W/Requirements of Later Edition of ASME Code.Approval to Use Applicable Portion of Later Edition Acceptable ML20247F2571989-09-0808 September 1989 SER Accepting Licensee Submittal in Compliance W/Atws Rule, 10CFR50.62, Requirements for Reduction of Risk from ATWS Events for Light-Water Cooled Nuclear Power Plants ML20246M6701989-08-29029 August 1989 Safety Evaluation Supporting Amend 56 to License NPF-38 ML20244D9781989-06-13013 June 1989 Supplemental Safety Evaluation Supporting Util Dcrdr Program That Satisfies Requirements of Suppl 1 to NUREG-0737 ML20244C8151989-06-0606 June 1989 Safety Evaluation Supporting Util First Interval Inservice Insp Program ML20244C8121989-06-0606 June 1989 Safety Evaluation Accepting Relief from Performing Inservice Insp Program Re Volumetric Exam on Inside Radius Section of Main Steam & Feedwater Nozzles ML20247J9431989-05-24024 May 1989 SER Accepting Util Response to Generic Ltr 83-28, Reactor Trip Sys Reliability ML20235B6381989-02-0707 February 1989 Safety Evaluation Re Inservice Testing Program & Requests for Relief from ASME Code,Section Xi.Program for Pumps & Valves & Request for Relief Acceptable.Relief Requests May Not Be Implemented W/O Prior NRC Approval ML20154R7301988-09-28028 September 1988 Safety Evaluation Supporting Util 840206 Response to Generic Ltr 83-28,Item 4.5.1, Reactor Trip Sys Reliability (Sys Functional Testing) ML20154N7071988-09-22022 September 1988 Safety Evaluation Accepting Util Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1 & 3.2.2 Re post-maint Test Program for safety-related Components ML20207F3391988-08-0909 August 1988 Safety Evaluation Supporting Proposed Rev to Tech Spec Bases B 3/4.7.6, Control Room Air Conditioning Sys ML20151M5281988-07-21021 July 1988 Safety Evaluation Re Control Sys Single Failure Study ML20155D6231988-05-27027 May 1988 Safety Evaluation Supporting Conditional Approval of Fracture Mechanics Analysis of Reactor Vessel Flaw Indications in Hot Leg Nozzle to Shell Weld for Util ML20154A9631988-05-0505 May 1988 Safety Evaluation Accepting Util 880303 Request for Partial Exemption from 10CFR50,App J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. Util May Continue Type a Testing When Excessive Leakage Identified ML20151B5691988-03-30030 March 1988 SER Accepting Util Proposal Re Item 2.2.1 of Generic Ltr 83-28 Concerning Equipment Classification Programs for All safety-related Components ML20237D7811987-12-21021 December 1987 Safety Evaluation Supporting Util Proposal Re Boraflex Surveillance Program in Spent Fuel Storage Racks ML20236X6491987-12-0101 December 1987 Safety Evaluation Accepting 871006 Request for Reduced Duration Integrated Leak Rate Test at Facility,Based on Methodology in BN-TOP-1,Rev 1, Testing Criteria for Integrated Leak Rate Testing of Primary Containment.. ML20236U1821987-11-24024 November 1987 Safety Evaluation Supporting Util 870724 Evaluation Demonstrating That Adequate Shoulder Gap Will Be Provided in Cycle 3 & Subsequent Cycles ML20236F7651987-10-27027 October 1987 Safety Evaluation Supporting Util 870626 & 0701 Surveillance Program & Results of Confirmatory Analyses Program.Util Satisfactorily Completed Confirmatory Analyses Program Demonstrating Adequacy of Basemat.Bnl Evaluation Rept Encl ML20236C0001987-10-20020 October 1987 Safety Evaluation Re Generic Ltr 83-28,Items 4.1,4.2.1 & 4.2.2 Concerning Preventive Maint Program for Reactor Trip Breakers/Maint & Trending.Licensee Position on Items Acceptable ML20237G7881987-08-20020 August 1987 Safety Evaluation Accepting Util 870731 Proposed Change to Bases Section of Tech Specs,Reflecting Commitment to 1982 Rev of ASTM E 185 Re Reactor Vessel Surveillance Program Required by 10CFR50,App H.Rev to Page B 3/4 4-7 Encl ML20249C8131987-07-21021 July 1987 Safety Evaluation Supporting Amend 20 to License NPF-38 ML20215E9781986-12-10010 December 1986 Safety Evaluation Supporting Addl Delay in Implementing Charcoal Filter Deluge Sys Mods Since Fire Protection Capability Provided ML20215B1881986-12-0808 December 1986 Safety Evaluation Re Util 860902 Submittal of CEN-335(c)-P, Waterford Unit 3,Cycle 2,Shoulder Gap Evaluation Rept, in Response to License Condition 2.c.7.Shoulder Gaps in All Fuel Acceptable Through Cycle 2 ML20211A2311986-05-29029 May 1986 Safety Evaluation Supporting Util 860123 & 0220 Responses to 10CFR50.61 Re Pressurized Thermal Shock Rule.Submittal of Reevaluation of Rt(Pts) & Comparison W/Predicted Value in Future pressure-temp Submittals Required ML20198C5541986-05-15015 May 1986 Safety Evaluation Accepting Util Response to Items 3.1.3 & 3.2.3 of Generic Ltr 83-28 Requiring Licensee Review of Existing Tech Specs for post-maint Testing Requirements That May Degrade Safety.Items Closed ML20197E3651986-05-0606 May 1986 Safety Evaluation Supporting Util Large Break LOCA ECCS Analysis ML20203Q0931986-04-22022 April 1986 Safety Evaluation Supporting Util 850613 & 860311 Responses Re Confirmatory Tests of Auxiliary Pressurizer Spray Sys. Design Satisfies Requirements of BTP Rsb 5-1 W/Single Failure of Charging Loop Isolation Value ML20137Z1641985-12-0202 December 1985 Safety Evaluation Supporting Release of Shift Advisors from Advisory Duties 1999-07-23
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217F2891999-10-13013 October 1999 Drill 99-08 Emergency Preparedness Exercise on 991013 ML20217G7211999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Waterford 3 Ses. with ML20211Q2141999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Waterord 3 Ses.With ML20210Q6361999-07-31031 July 1999 Corrected Monthly Operating Rept for July 1999 for Waterford 3 ML20210S0581999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Waterford 3.With ML20210D8951999-07-23023 July 1999 Safety Evaluation Accepting First 10-yr Interval Inservice Insp Plan Requests for Relief ISI-018 - ISI-020 ML20209H3781999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Waterford 3 Ses. with ML20195J8951999-06-17017 June 1999 Safety Evaluation Granting Relief for Listed ISI Parts for Current Interval,Per 10CFR50.55a(g)(5)(iii) ML20207E8631999-06-0303 June 1999 Safety Evaluation Accepting Licensee 990114 Submittal of one-time Request for Relief from ASME B&PV Code IST Requirements for Pressure Safety Valves at Plant,Unit 3 ML20195K3391999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Waterford 3 Ses.With ML20195C3041999-05-28028 May 1999 Annual Rept on Abb CE ECCS Performance Evaluation Models ML20206S7401999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Waterford 3.With ML20205T2621999-04-22022 April 1999 LER 99-S02-00:on 990216,contract Employee Inappropriately Granted Unescorted Access to Plant Protected Area.Caused by Personnel Error.Security Personnel Performed Review of Work & Work Area That Individual Was Involved with ML20205N9671999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Waterford 3 Ses.With ML20205E8531999-03-30030 March 1999 Corrected Pages COLR 3/4 1-4 & COLR 3/4 2-6 to Rev 1, Cycle 10, Colr ML20205A6331999-03-25025 March 1999 SER Accepting Request to Use Mechanical Nozzle Seal Assemblies as an Alternative Repair Method,Per 10CFR50.55a(a)(3)(i) for Reactor Coolant Sys Applications at Plant,Unit 3 ML20204H1401999-03-23023 March 1999 Rev 1 to Engineering Rept C-NOME-ER-0120, Design Evaluation of Various Applications at Waterford Unit 3 ML20204H1231999-03-22022 March 1999 Rev 1 to Design Rept C-PENG-DR-006, Addendum to Cenc Rept 1444 Analytical Rept for Waterford Unit 3 Piping ML20204H2451999-03-22022 March 1999 Rev 2 to C-NOME-SP-0067, Design Specification for Mechanical Nozzle Seal Assembly (Mnsa) Waterford Unit 3 ML20204F0791999-03-17017 March 1999 Rev 1 to Waterford 3 COLR for Cycle 10 ML20207M9231999-03-12012 March 1999 Amended Part 21 Rept Re Cooper-Bessemer Ksv EDG Power Piston Failure.Total of 198 or More Pistons Have Been Measured at Seven Different Sites.All Potentially Defective Pistons Have Been Removed from Svc Based on Encl Results ML20207F3491999-03-0505 March 1999 LER 99-S01-00:on 990203,contraband Was Discovered in Plant Protected Area.Bottle Was Determined to Have Been There Since Original Plant Construction.Bottle Was Removed & Security Personnel Performed Search of Area.With ML20204B5141999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Waterford 3.With ML20203H8591999-02-17017 February 1999 Safety Evaluation Accepting Licensee Second Ten Year ISI Program & Associated Relief Requests for Plant,Unit 3 ML20199H6261999-01-21021 January 1999 Safety Evaluation Accepting Classification of Instrument Air Tubing & Components for Safety Related Valve Top Works.Staff Recommends That EOI Revise Licensing Basis to Permit Incorporation of Change ML20199C9101998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Waterford 3.With ML20196F4911998-12-0101 December 1998 SER Accepting Request for Relief ISI2-09 for Waterford Steam Electric Station,Unit 3 & Arkansas Nuclear One,Unit 2 ML20206N4131998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Waterford 3.With ML20195C4841998-11-0606 November 1998 SER Accepting QA Program Change to Consolidate Four Existing QA Programs for Arkansas Nuclear One,Grand Gulf Nuclear Station,River Bend Station & Waterford 3 Steam Electric Station Into Single QA Program ML20195E5161998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Waterford 3.With ML20155C1351998-10-26026 October 1998 Rev B to Entergy QA Program Manual ML20154K0801998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Waterford 3 Ses. with ML20151W8331998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Waterford,Unit 3. with ML20237B6831998-08-17017 August 1998 LER 98-S01-00:on 980723,discovered That Waterford 3 Physical Security Plan,Safeguards Document Was Not Under Positive Control of Authorized Person at All Times.Caused by Human Error/Inappropriate Action.Counseled Employee Involved ML20237C5661998-08-17017 August 1998 Safety Evaluation Accepting Licensee Request for Exemption from Section Iii.O of Appendix R to 10CFR50 ML18066A2771998-08-13013 August 1998 Part 21 Rept Re Deficiency in CE Current Screening Methodology for Determining Limiting Fuel Assembly for Detailed PWR thermal-hydraulic Sa.Evaluations Were Performed for Affected Plants to Determine Effect of Deficiency ML20237B5261998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Waterford 3 ML20236S9031998-07-22022 July 1998 SER Accepting Rev 19 to Quality Assurance Program for Waterford Steam Electric Station,Unit 3 ML20198H3911998-07-14014 July 1998 Non-proprietary Rev 5 to HI-961586, Thermal-Hydraulic Analysis of Waterford-3 Spent Fuel Pool ML20236N4181998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Waterford,Unit 3 ML20248E7781998-06-0101 June 1998 Annual Rept on Abb CE ECCS Performance Evaluation Models ML20249A4711998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Waterford 3 Ses ML20196A4051998-05-31031 May 1998 Rept of Facility Changes,Tests & Experiments,Per 10CFR50.59 for 970601-980531. with ML20198H4681998-05-20020 May 1998 Non-proprietary Rev 1 to HI-981942, Independent Review of Waterford Unit 3 Spent Fuel Pool Cfd Model ML20247A3891998-05-0101 May 1998 SG Eddy Current Examination (8th Refueling Outage) ML20247F6761998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Waterford,Unit 3.W/ ML20217M8951998-04-30030 April 1998 QA Program Manual ML20217P8281998-04-0707 April 1998 Safety Evaluation Accepting Relief Authorization for Alternative to Requirements of ASME Section Xi,Subarticle IWA-5250 Bolting Exam for Plants,Per 10CFR50.55a(a)(3)(i) ML20216B1751998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Waterford 3 Ses ML20217M1411998-03-0303 March 1998 Rev 2 of Waterford 3 Cycle 9 Colr 1999-09-30
[Table view] |
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[ T NUCLEAR REGULATORY COMMISSION
$ j WASHINGTON, D. C. 20555 e
9 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FACILITY OPERATING LICENSE N0. NPF-38 LOUISIANA POWEP AND LIGHT COMPANY WATERF0PD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382 INTRODUCTION During the 1988 refueling cuttge the Louisiana Power and Light Company (the licensee) perforrred an inservice inspection of several welds in the reactor pressure vessel in the Waterford Unit 3 (Waterford). The examinations were performed in accordance with the 1980 Edition through Winter 1981 Addenda of ASME Section XI. As a result of conventional ultrasonic testing of the welds three recordable indications were observed. That examination indicated that two flaws were less than the allowable size criteria in Article IWB-3500 of ASME Section XI and one indication exceeded the criteria. The licensee again performed the examination of the flaws using equipment to enhance the character-izatica and as a result, all three flaws exceeded the criteria. Article IWB-3600 esta-blishes rules for evaluating flaw indications that exceed the acceptance criteria in Article IWB-3500. Paragraph IWB-3610 states that the evaluation procedures shall be the responsibility of the Owner and chall be subject to approval by the regulatory authority having jurisdiction at the plant site. In a letter dated May 16, 1988 the licensee submitted for staff review the technical evaluation of the subject flaws.
STAFF EVALUATION A. Nondestructive Examination During the inservice ultrasonic examination of the hot leg nozzle-to-shell weld No. 01 - 021 three recordable indications were observed. Two (2) of the indications were detected with a 0 degree 2.25 mHz longitudinal wave examination from the nozzle bore, and the remaining indication was detected with a 20 degree, 2.25 mHz longitudinal wave examination from the nozzle bore. These indications are located within the weld at or near the weld /
nozzle forging fusion line. The O degree longitudinal wave indications were determined to meet the acceptance standards in Table IWB-3512-1 of the ASME Code Section XI, 1980 Edition through the 1981 Addenda, while the 20 degree longitudinal wave indication exceeded the allowable limits of Table IWB-3512-1.
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In an effort to further characterize these indications, particularly the 20 degree longitudinal wave indication, supplemental examinations were per-formed using the Dynacon Ultrasonic Data Recording and Processing System (UDRPS). Based on the UDRPS examination results all three reflectors ex-ceeded the allowable limits of Table IWB-3512-1.
The upper shell thickness at the location of the indications is 10.75 inches.
The ultrasonic examination indicates that the three flaws are located 2.73, 3.34 and 4.01 inches from the reactor vessels outside surface. Hence all the flaws may be considered embedded and are located closer to the outside surface.
The straight beam ultrasonic techniques produced relatively strong reflec-tions. Satellite pulses were observed with the supplemental characteriza-tion technique suggesting that the flaw indications originate from volumetric type defects, such as slag er porosity. Even with small flaws the estimated size is more consistent with the beam size of the transducer rather than the size of the flaw (for beam sizes greater than the size of the flaw).
The staff has reviewed the examination results and concludes that the licensee's dimensions of the flaw indications are conservative. The conven-tional data indicates that the largest flaw has a depth of 1.19 inch. For the largest flaw, the UDRPS examiration produced two separate reflectors which when sized according to Section XI proximity rules resulted in a depth of approximately 2.5 inches. The staff does not believe that hot leg nozzle-to-shell weld contains slag indications which have a depth of 2.5 inches. I The examination data suggests that the reported dimensions are a function of the characteristics of the transducer rather than a measurement of the size of the flaw indication.
B. Fracture Mechanics Evaluation l
The most conservative dimensions obtained during the ultrasonic examinations l were used in the fracture mechanics evaluation of the hot leg nozzle-to-shell weld. The licensee has provided a flaw evaluation chart for the Waterford 3 hot leg nozzle-to-shell weld. The method and criteria used in the fracture mechanics analyses are documented in Reference 1. The portions of this document that were related to the Waterford flaw evaluation were documented in Enclosure 1 of licensee's May 16, 1988 submittal. The frac- I ture mechanics analyses that were performed to develop the flaw evaluation chart were in accordance with the methodology and criteria specified in Article IWB-3600 and Appendix A of the ASME Section XI except that stresses were not linearized and stress intensity factors were not calculated in accordance with the recomendations in Appendix A. In lieu of linearizing the stress, the method used represented the actual stress profile by a third order polynomial. Stress intensity factors were calculated using the ex-pressions of Reference 2. These stress intensity factor expressions have been shown to be applicable to vessels in Reference 3. These stress profiles and stress intensity factor expressions provide a more accurate determination of the critical flaw size, and are particularly important during the evalua-tion of emergency and faulted conditions where the stress profile is generally nonlinear and often very steep.
Important parameters in a fracture mechanics analyses are the materials' brittle fracture resistance and the projected flaw growth rate during opera-tion of the component. The standard measurement of the brittie fracture resistance for the Waterford reactor vessel material are. their crack initia-tion and arrest fracture toughness. These values of fracture tcughness are used to detemine a critical flaw size. Vestinghouse indicates that the critical flaw size calculation used the crack initiation and arrest fracture tcughness for vessel materials that are recomstended in Appendix A of the ASME Section XI. The critical flaw size for the hot leg nozzle-to-shell of O'F weldlocationwasdeterminedusingarefergncetemperature,RTThese and an upper shelf toughness of 200ksi(in) . N values are S this location in the reactor vessel because the materials in this location are not subject to significant amounts of neutron irradiation and the RT valuewasthehighestvalueforallmateriallocatedinthehotlegnozzYST -
to-shell region. The reference temperature for all materials in the hot leg i nozzle-to-shell region are reported on FSAR Table 5.2.-6. The staff reviewed
{
this data in NUREG-0708 Supplement No. 1, "Safety Evaluation Report" l October 1981. This staff evaluation indicates that a reference temperature l of 0 F should be used for all welds outside the Waterford beltline region.
The amount of projected flaw growth was determined to be negligible. The calculation was performed for the reactor vessel design transients that arc listed in Table 2-1 in Enc 1csure 1 of the April 28, 1988 submittal. The rate of fatigue growth was calculated using the ASPE reference curve for air )
environment. Since the flaws under evaluation are embedded, this method of calculating the flaw growth rate is acceptable.
The flaw evaluation chart was constructed from fracture mechanics analyses I cf reactor vessel design and operating transients, that are listed in Table 2-1. These transients included events during upset, test and emergency and faulted conditions.
l The evaluation of emergency and faulted conditions included pressurized themal shock (PTS) events which were categorized in Reference 4. Reference 4 was a PTS evaluation of Calvert Cliffs. The Calvert Cliffs plant was chosen as the representative generic Combustion Engineering designed plant for the pressurized themal shock issue. Since Waterford is a Combustion Engineering designed plant, the transients in Reference 4 were assumed to be representative of events for the Waterford plant. The PTS transients in-cluded moderate to severe cooldowns with sever repressurizations up to the relief valve setting of 2500 psi. Since severe cooldowns will produce compressive themal stresses on flaws located near the outside surface, the flaws in the Waterford nozzle-to-shell weld which are located near the out-side surface may be conservatively evaluated by neglecting the compressive themal stress and considering the maximum pressure during the event. Hence, the limiting PTS transients for this evaluation were events involving re-pressurization to the relief valve set point, which included large steam line and small steam line breaks.
After considering all events during upset, test, emergency and faulted conditions, the limiting event for the hot leg nozzle-to-shell location was determined to be the Primary Side Hydrotest with pressurization to 3105 psi.
The flaw evaluation chart resulting from the fracture mechanics analysis of this event indications that the reported flaw sizes meet the criteria in Article IVB-3600 for the 40 year service life of the plant.
In addition to tne reactor vessel design transients, which are listed in Tables 2-1 and 4-1, the licensee evaluated a postulated low temperature overpressure (LTOP) event, which was not mitigated by the LTOP protection system. To determine whether this event was either an upset or an emergency /
faulted condition the licensee performed a probabilistic risk assessment.
The assessment included failure probabilities for the valves in the Waterford LTOP system. The valve failure probabilities form the basis for the risk assessment. The failure probabilities used in the licensees assessment com-pare favorably with the valve failure rates identified in NUREG/CR 2728 "Interim Reliability Evaluation Procedure Guide," January 1983. The licen-see's detailed probabilistic risk assessment indicated that this event should be classified as a faulted condition. The licensee's analysis of the postu-lated LTOP event indicates that the LTOP event is not a governing transient because it is much less severe than the other faulted conditions.
The NRC required licensees to install LTOP protection systems in 1979. Since the industry installed LTOP protection systems, there has been only one event in which the LTOP system did not mitigate the event. This event occurred on November 28, 1981 atTurkeyPointUnit4(Reference 5). In this event, the pressure rose to 1100 psi, at a temperature of 110 F. Pressurized water reactors (PWRs) have accumulated approximately 400 years of plant cperation since installation of LTOP protection systems. Since only one event has occurred in 300 be expected to occur during the 40 year life of a PWR nuclear power plant. Hence, according to Appendix A, 10 CFR 50, the event is not an anticipated operational occurrence and may be considered an emergency / faulted condition.
To conservatively bound LT0P events for the Waterford reactor vessel, the staff has performed a fracture mechanics analysis for the Waterford reactor vessel in which the postulated event occurred at 110*F and pressurized the vessel at 1500 psi. The analysis was performed using the nethodology des-cribed in Appendix A of ASME Section XI. The staff's evaluation indicates that for the postulated event, the flaws in the nozzle-to-shell weld will meet the acceptance criteria in Article IWB-3600 for emergency / faulted conditions.
CONCLUSIONS
- 1) Based on the licensee's and the staff's independent evaluation of a postulated LTOP event, the flaws in the hot leg nozzle-to-shell weld No. 01 - 021 satisfy the analytical evaluation criteria in Article IWB-3600. Based on these analyses, the flaws in the weld will not grow during the life of the plant to a size that will affect the integrity of the reactor vessel. The reactor vessel is acceptable for the 40 years of service life of the plant.
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- 2) However, the flaws in the hot leg nozzle-to-shell weld are conditionally acceptable. Pursuant to ASME Section XI paragraphs IWB-3122.4(b) and IWB-2420(b), weld No. 21 will be reexamined during the next three inspection periods. The staff concludes that the licensee should evaluate the use of an additional transducer with a narrower beam spread for the reexamination.
A comparison of the results with the transducer used during the 1988 l examination and another with optimum characteristics at the location of the flaw should provide a better definition of the dimensions of the reflector.
References
- 1. Bamford, W.H., et. al., Handbook on Flaw Evaluation Waterford Unit 3 Reactor Vessel Outlet Nozzle-To-Shell Welds," May 1988.
- 2. Shah, R.C. and Kobayashi, A.S., "Stress Intensity Factor for an Elliptical Crack Under Arbitrary Loading", Engineering Fracture Mechanics, Vol. 3, 1981, pp. 71-96.
- 3. Lee, Y.S. and Bamford, W.H., "Stress Intensity Factor Solutions for a Lengitudinal Buried Elliptical Flaw in a Cylinder Under Arbitrary Loads",
presented at ASME Pressure Vessel and Piping Conference, Portland Oregon, June 1983. Paper 83-PVP-92.
- 4. Shelby, D.L. et al., Pressurized Themal Shock Evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant," Oak Ridge National Labs Report ORNL/
TM 9408, NUREG-CR 4022, September 1985.
- 5. W.D. Lanning, "Low Temperature Overpressure Event at Turkey Point Unit 4."
Case Study Report by Office for Analysis and Evaluation of Operational Data, NRC, March 1984.
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