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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
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7 4
i I[atRicyogg UNITED STATES NUCLEAR REGULATORY COMMISSION g ;E WASHINGTON, D. C. 20555
% d 4.....s SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION BORAFLEX SURVEILLANCE PROGRAM FACILITY OPERATING LICENSE NO. NPF-38 LOUISIANA POWER AND LIGHT COMPANY WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET N0. 50-382
1.0 INTRODUCTION
By letter dated September 16 and December 15, 1987, Louisiana Power and !
Light Company (LP&L, the licensee) requested NRC staff concurrence of I LP&L's proposal to. revise their commitment for surveillance of Boraflex in the spent fuel storage racks (SFSRs) at Waterford 3, 2.0 DISCUSSION In its FSAR response to NRC Question No. 282.4, dated April 1981, LP&L committed to conduct surveillance testing on samples of the Boraflex poison material (coupons) in the Waterford 3 SFSRs. The surveillance coupons are two-inch-square (0.1 inch thick) pieces of Boraflex enclosed in stainless steel, which are contained within a Boraflex train assembly in the SFSR. The surveillance was to be performed over a five year period, beginning with the second refueling outage, and includes measure- l ments of dimensions, hardness, weight, boron content through chemical ;
analysis, and the neutron attenuation of the Boraflex surveillance coupons l which were exposed to discharged fuel assemblies of greatest burnup. l These measurements were intended to demonstrate (1) that there is sufficient )
boron in the Boraflex to assure subcritical conditions in the SFSRs as !
assumed in the licensing analysis, and (2) that the Boraflex material does I not degrade to an unacceptable condition under gamma irradiation from )
spent fuel assemblies.
The purpose of the Boron-10 in the material is to reduce the neutron multiplication in the SFSR by absorbing neutrons. ;
Brand Industrial Services Inc. (BISCO), the Boraflex manufacturer, has determined that the B-10 density in Boraflex is higher than the value assumed in the criticality analysis. Furthermore, the analysis of coupon data taken prior to irradiation which was performed by System Services Inc , shows that the lowest B-10 density in the Boraflex coupons is 29% l higher than that used in the criticality analyses. Thus, the criticality analysis is conservative with respect to boron content in assuring a subcritical condition in the SFSRs.
8712240181 871221 PDR P ADOCK 05000382 I PDR l
The Boraflex poison material may lose its B-10 content by neutron absorption or by gamma radiation induced physical degradation. Boron depletion due to neutron absorption in the Boraflex is not a significant concern because the low neutron flux in.the pool produces only a small number of neutron absorption interactions. However, Boraflex physical; degradation due.to gamma irradiation is a potential contributor to B-10 losses in the poison material. As long as the material is not physically degraded the boron content will not' change. The required chemical analysis to determine the actual boron content in the Boraflex coupons (the original commitment) is thus not necessary, and for the same reason, the neutron attenuation analysis is also unnecessary. Boraflex neutron attenuation capability remains unchanged as long as the material has not suffered any physical degradation due to gamma irradiation.
The original ~ commitment mentions the parameters that should be measured i (thickness, length, width, weight, and hardness) but it does not describe how or when (based on these parameters) the Boraflex coupons will be considered to be unacceptably physically degraded. The technical basis for.this commitment needs to be improved to reflect new industry developments.
Experience has shown that the surveillance coupons are not representative of the full length Boraflex insert. Wisconsin Electric Power Company ;
(WEPCO) performed a Boraflex coupon surveillance similar to the original J LP&L commitment (the design and manufacturer of the spent fuel storage J racks as well as the Boraflex material at Point Beach, WEPC0's nuclear
-facility are the same as for Waterford 3), and found that the Boraflex ,
coupons were not representative of the full length inserts. When comparing I the full length inserts with coupons exposed to 1 E+10 rads (equivalent ]
to 20 year gamma dose exposure in the SFSRs), WEPC0 found that the full- 1 length inserts did not show any significant signs of degradation. However, !
the coupons showed decreases in thickness, width, length, and weight, and I they were fragile and easily broken.
LP&L has determined that a modification to the existing commitment is appropriate. The bases for this modification are: (1) the need to improve the technical basis to support the required analysis, (2) the surveillance coupons are not representative of the Boraflex full-length inserts, and (3) there is not enough information about Boraflex integrity in the industry at this time in order to set up an appropriate surveillance program.
As an alternative Boraflex Surveillance Program, LP&L proposes the j following revised commitments:
- 1. The current Boraflex coupon surveillance commitment will not be i performed.
- 2. LP&L will develop a log to track the gamma dose buildup in the spent fuel storage racks.
b !
L i
l
1
- 3. LP&L will keep abreast of new industry developments on Boraflex integrity for the next few-years. An appropriate surveillance program will be proposed to the NRC by January 1, 1993.
l
~ 3. 0 EVALUATION BISCO has reported that no evidence of Boraflex deterioration has been I found through cumulative gamma radiation in an excess of 1 E+11 rads. I Also, at the present time BISCO is engaged in further studies, and they plan to conduct analysis of larger Boraflex samples under conditions which are similar to-those at'the Waterford 3 SFSRs. The results from 1 these studies are scheduled to be available at the end of'1988. Assessing the results from these studies and other pertinent industry experience is important prior to establishing a revised Boraflex surveillance program with a firm technical basis.
Furthermore, WEPC0 concluded that a full-length insert which received an acceleratedradiationdoseof1E+10radsgamma"hadgoodintegritywith no pieces missing, no cracking, or other degradation observed. Overall, the poison' insert "although brittle, had good integrity with minimal ]
degradation." -]
LP&L has estimated the gamma dose of a poison insert'in the Waterford 3 SFSR exposed to a fuel assembly with burnup'of 45 GWD/MTV, and three different power histories (low, average, and high). The calculations assumed also.that the Boraflex material was exposed to four similar spent j fuelaassemblies to maximize the dose. A 14-year period was assumed because that is the approximate time-it would take to fill-the Waterford 3 SFSRs while.still allowing for a full core discharge. The results on dose buildup from the assembly with high power history show that the estimated maximum dose that the Boraflex material will be exposed to in 14 years is 8.82 E+8 rads gamma (3.82% of WEPC0 dose).
On September 8, 1987, the NRC issued Information Notice 87-43 to alert recipients.to'a potentially significant problem pertaining to gaps iden-tified in the neutron absorber component, Boraflex, of the high-density spent' fuel storage racks at Quad Cities Unit 1. Shrinkage of the Boraflex sheet is expected to occur as the material is irradiated and is produced by two radiation-induced mechanisms, crosslinking and scissioning.
However, a contributor to the formation of these gaps is the local tensile stress created by the application of an adhesive compound (Dow Silicone
- 999) during the fabrication process of the Quad Cities' SFSRs.
This adhesive was applied to approximately the center of the stainless steel sheet in a discontinuous bead along the entire length. The bead was spread out to a width of approximately 2 "-3" with a stainless steel l scraper. The Boraflex was then rolled into place and pressed against the L stainless steel sheets. There were no specific procedures for this l process since the only intended function of the adhesive was to hold the
l l
l Boraflex in place during the SFSR fabrication process. The bonding i between the stainless steel and the Boraflex (created by the adhesive 1' compound), produces high local stresses in the sheet as the material shrinks, forming small gaps along the length of the sheet.
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Although the design used for the Waterford 3 SFSRs differs from the Quad Cities Units 1 and 2 design, uncertainties exist as to the effect of rack j design and manufacturing methods on shrinkage-induced gap formation. i Therefore, a surveillance program is needed to assure suitability of j Boraflex neutron absorber material for the life of the high density a storage racks. Based on the results of EPRI and BISCO programs to improve j understanding of the Point Beach Units 1 and 2 and Quad Cities Unit 1 anomolies (non-representative coupons and gaps), the licensee proposes to submit a surveillance program by January 1, 1993. In the next five years, i
the maximum gstimated dose that the Boraflex will be exposed to is estimated to be 7.3x10 Rads gamma. At this exposure, the amount of shrinkage is :
not expected to produce gaps that would significantly degrade the Boraflex neutron absorption performance. Therefore, a commitment to provide a d surveillance program by January 1, 1993 is acceptable. The surveillance program should include non-destructive examination of multiple, represen- { '
tative, full-length Boraflex panels that have received the maximum spent fuel pool exposure to examine for the presence of gaps. The full-length panels should be examined periodically, on a time frame consistent with increasing degrees of predicted shrinkage. Alternatively, an analysis based on the actual Waterford 3 configuration and data which demonstrates the effectiveness of the Boraflex poison material for increased degrees i of predicted shrinkage may be performed. The proposed program should incorporate actual data from the examination of approximately 10 representative full-length panels for the presence of gaps to provide a basis for the extended surveillance program.
4.0 CONCLUSION
j Based on the above evaluation, the staff finds the licensee's Boraflex surveillance program proposal acceptable. This includes the following revised commitments: l 1- The original Boraflex coupon surveillance program will not be performed.
2- LP&L will develop a log to track the gamma dose buildup in the spent fuel storage racks.
3- LP&L will keep abreast of new industry developments on Boraflex integrity for the next few years. An appropriate program will be proposed to the NRC by January 1, 1993.
, This surveillance program should include:
- data from the non-destructive examination of approximately 10 rept esentative full-length Boraflex panels for the presence of gaps; and l
L ___ __ _ __ .
. 1 periodic re-examination of multiple, representative, full-length Boraflex panels on a time frame consistent with increasing degrees of predicted shrinkage or, alternatively, an analysis based on the actual Waterford 3 configuration ar.d data which demonstrates the effectiveness of the Boraflex poison material for increased degrees of predicted shrinkage.
I Dated: December 21, 1987 Principal contributor: J. Wilson i
)