Safety Evaluation Re Util 861027 Request for Relief from Exam Requirements of Section XI of ASME Boiler & Pressure Vessel Code for Shell & Nozzle Welds in Regenerative Hxs. Request GrantedML20213G580 |
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Point Beach |
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Issue date: |
05/07/1987 |
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From: |
Office of Nuclear Reactor Regulation |
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To: |
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Shared Package |
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ML20213G556 |
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References |
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NUDOCS 8705180353 |
Download: ML20213G580 (6) |
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20212D5961999-09-15015 September 1999 Safety Evaluation Supporting Licensee IPEEE Process.Plant Has Met Intent of Suppl 4 to GL 88-20 ML20196J4251999-06-30030 June 1999 Safety Evaluation Authorizing Proposed Alternatives Described in Relief Requests VRR-01,ROJ-16,PRR-01 & VRR-02 ML20207D6751999-02-22022 February 1999 Assessment of Design Info on Piping Restraints for Point Beach Nuclear Plant,Units 1 & 2.Staff Concludes That Licensee Unable to Retrieve Original Analyses That May Have Been Performed to Justify Removal of Shim Collars ML20206R9001999-01-13013 January 1999 SER Accepting Nuclear Quality Assurance Program Changes for Point Beach Nuclear Plant,Units 1 & 2 ML20198C7671998-12-10010 December 1998 Safety Evaluation Accepting Licensee Proposed Alternative to ASME BPV Code,1986 Edition,Section XI Requirement IWA-2232, to Use Performance Demonstration Initiative Program During RPV Third 10-yr ISI for Plant,Unit 2 ML20236Q3161998-07-10010 July 1998 Safety Evaluation Accepting Licensee Proposed Alternative to ASME Code Requirements PTP-3-01 & PTP-3-02 ML20236L6771998-07-0707 July 1998 Safety Evaluation Approving Wepco Implementation Program to Resolve USI A-46 at Point Beach NPP Units 1 & 2 ML20217F3131998-04-17017 April 1998 Safety Evaluation Accepting Proposed Alternative to ASME Code for Surface Exam of Nonstructural Seal Welds,For Plant, Unit 1 ML20217A8501998-03-19019 March 1998 SER Accepting Proposed Changes Submitted on 980226 by Wiep to Pbnp Final SAR Section 1.8 Which Will Impact Commitments Made in Pbnp QA Program Description.Changes Concern Approval Authority for Procedures & Interviewing Authority ML20216J0101998-03-17017 March 1998 Safety Evaluation Accepting Third 10-yr Inservice Insp Interval Relief Request RR-1-18 for Plant ML20198L1151998-01-0808 January 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Point Beach Nuclear Plant,Units 1 & 2 ML20198L4671998-01-0202 January 1998 SER Approving Request for Relief VRR-4B to Inservice Testing Program Wisconsin Electric Power Co,Point Beach Nuclear Plant,Units 1 & 2 ML20197J9341997-12-12012 December 1997 Safety Evaluation Accepting Licensee Request for Relief from Performing Inservice Volmetric Exam of Inaccessible Portions of RPV Lower Shell to Lower Head Ring Weld During 10-yr ISI Interval of Plant,Unit 2 ML20137U4991997-04-10010 April 1997 Safety Evaluation Accepting Proposed Alternatives Contained in Requests for Relief RR-1-17 & RR-2-21 ML20129G6901996-10-0303 October 1996 SER Accepting Request for Relief from ASME Code Repair Requirements for ASME Code Class Three Piping at Plant ML20062J4991993-10-28028 October 1993 Safety Evaluation Granting IST Relief Requests Per 10CFR50.55a(a)(3)(ii) & 10CFR50.55a(f)(4)(iv) ML20062F1361990-09-25025 September 1990 SE Accepting Util Responses to Generic Ltr 83-28,Item 1.2, Post-Trip Review - Data & Info Capability ML20248A0101989-09-18018 September 1989 Safety Evaluation Re Containment Liner Leak Chase Channel Venting.Concurs W/Licensee That Plant Does Not Need to Vent Containment Liner Weld Leak Chase Channels During Test ML20246H0121989-07-0707 July 1989 Safety Evaluation Accepting Util 880325 & 1117 Responses to NRC Bulletin 88-002, Rapidly Propagating Fatigue Cracks in Steam Generator Tubes ML20245B0311989-06-14014 June 1989 Safety Evaluation Supporting Util Response to Generic Ltr 83-28,Item 4.5.3 Re on-line Functional Testing of Reactor Trip Sys.Existing Intervals for on-line Functional Testing Consistent W/High Reactor Trip Sys Availability ML20207E4191988-08-0404 August 1988 Safety Evaluation Supporting Compliance W/Atws Rule 10CFR50.62, Requirements for Reduction of Risk from ATWS Events for Light Water Cooled Nuclear Power Plants ML20151R6771988-08-0202 August 1988 Safety Evaluation Granting Request for Relief from ASME Code,Section XI Evaluation Requirements ML20151N2191988-07-27027 July 1988 Safety Evaluation Supporting Util Proposal Re Design of Switchgear Room,Per Sections Iii.G & Iii.L of App R to 10CFR50 ML20150C1311988-06-21021 June 1988 Safety Evaluation Accepting Responses to Generic Ltr 83-28, Item 2.1,confirming That Program Exists for Identifying, Classifying & Treating Components Required for Performance of Reactor Trip Function as safety-related ML20154H5791988-05-12012 May 1988 Safety Evaluation Supporting Conclusions That Rev 1 to Offsite Dose Calculation Manual (ODCM) Uses Methods Consistent W/Staff Requirements,However Some Discrepancies Identified.Odcm & Environ Manual Should Be Revised ML20148H4551988-03-24024 March 1988 Safety Evaluation Accepting Util 840405 Response to Generic Ltr 83-28,Item 2.1,(Part 2) Re Vendor Interface Programs & Reactor Trip Sys Components ML20235K9241987-07-0909 July 1987 Safety Evaluation Re Reactor Pressure Vessel Flaw.Flaw Conditionally Acceptable Per Subarticle IWB-3123 of Section XI of ASME Code & Therefore Requires Augmented Inservice Insps Based on 10CFR50.55(g)(4) ML20213G5801987-05-0707 May 1987 Safety Evaluation Re Util 861027 Request for Relief from Exam Requirements of Section XI of ASME Boiler & Pressure Vessel Code for Shell & Nozzle Welds in Regenerative Hxs. Request Granted ML20206K6011987-04-10010 April 1987 SER Supporting Util 860513 Proposed Replacement of Hydraulic Snubbers W/Energy Absorbers on Main Steam Bypass Line ML20210P2781987-02-0505 February 1987 Safety Evaluation Supporting Util 831107 & 860411 Responses to Generic Ltr 83-28,Item 4.5.2 Re Reactor Trip Sys Reliability on-line Testing.Plant Designed to Permit on-line Functional Testing of Diverse Trip Features of Breakers ML20214U6081986-11-26026 November 1986 Safety Evaluation Supporting Util 850516 Capsule T Summary Rept Re Use of Reactor Vessel Pressure Temp Limits Specified in Figures 15.3.1-1 & 15.3.1-2 of Tech Specs.Temp Limits Valid & May Continue to Be Used ML20206S7091986-09-16016 September 1986 Safety Evaluation on Util 850426 Response to Open Items Re Generic Ltr 81-14, Seismic Qualification of Auxiliary Feedwater Sys (Afws). Reasonable Assurance Exists That Afws Will Perform Required Safety Function Following SSE ML20214L9311986-09-0404 September 1986 Corrected Safety Evaluation Re Projected Values of Matl Properties for Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.Licensee Projections Acceptable ML20207D6781986-07-11011 July 1986 Safety Evaluation Accepting Util Responses to Generic Ltr 82-33 Re post-accident Monitoring Instrumentation Compliance W/Guidelines of Reg Guide 1.97,Rev 2,subj to Listed Condition.Portions of Rev 1 to EGG-EA-6771 Encl ML20138N7801985-10-31031 October 1985 Safety Evaluation Granting Util 840706 Relief Requests for Second 10-yr Inservice Insp Interval.Review of Requests for Relief from ASME Code Section XI Requirements Summarized in Encl Tables ML20134A4821985-10-24024 October 1985 Safety Evaluation Supporting Util Response to Generic Ltr 83-28,Items 3.1.1,3.1.2,4.1 & 4.5.1 Re post-maint Testing (Reactor Trip Sys Components) & Reactor Trip Sys Reliability.Programs Outlined in Acceptable ML20134A6051985-10-22022 October 1985 Safety Evaluation Re Util 831107 & 850910 Responses to Generic Ltr 83-28,Item 1.1, Post-Trip Review Program Description & Procedures. Program & Procedures Acceptable ML20138H1721985-10-18018 October 1985 Safety Evaluation Accepting Util 831107 Response to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Re post-maint Testing ML20133G4171985-07-29029 July 1985 Safety Evaluation Accepting Util 831108 Response to Generic Ltr 83-28,Item 1.1 Re post-trip Review.Response to Listed Deficiencies,Including Development of Systematic Safety Assessment Program for Unscheduled Reactor Trips Required ML20129H7871985-05-16016 May 1985 Safety Evaluation Supporting Licensee Response to Generic Ltr 83-28,Items 4.2.1 & 4.2.2 Re Reactor Trip Sys Reliability,Provided Corrective Action Taken If Higher than Normal Valves Observed in Trip Force & Response Time Values ML20205H2171984-09-10010 September 1984 Supplemental Safety Evaluation Re Util 820820 & 860113 Requests for Relief from Inservice Insp Requirements. Volumetric Exam Acceptable Method for Detecting O.D. Initiated Flaws.Relief from Surface Exams Should Be Granted ML20204F5381983-04-25025 April 1983 Safety Evaluation of Util Preferred Ac Power Sys Conformance GDC 17.Proximity of Low Voltage Transformers Does Not Fully Meet GDC 17 Requirements for Physical Separation,But Deluge Sprinkler Sys Adequate 1999-09-15
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARNPL-99-0569, Monthly Operating Repts for Sept 1999 for Pbnp,Units 1 & 2. with1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Pbnp,Units 1 & 2. with ML20212D5961999-09-15015 September 1999 Safety Evaluation Supporting Licensee IPEEE Process.Plant Has Met Intent of Suppl 4 to GL 88-20 NPL-99-0051, Monthly Operating Repts for Aug 1999 for Pbnp,Units 1 & 2. with1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Pbnp,Units 1 & 2. with NPL-99-0449, Monthly Operating Repts for July 1999 for Pbnp,Units 1 & 2. with1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Pbnp,Units 1 & 2. with ML20196J4251999-06-30030 June 1999 Safety Evaluation Authorizing Proposed Alternatives Described in Relief Requests VRR-01,ROJ-16,PRR-01 & VRR-02 ML20209D2691999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Pbnps,Units 1 & 2 ML20196F3341999-06-22022 June 1999 Safety Evaluation for Implementation of 422V+ Fuel Assemblies at Pbnp Units 1 & 2 ML20195F9781999-06-10010 June 1999 Unit 2 Refueling 23 Inservice Insp Summary Rept for Form NIS-1 ML20209D2751999-05-31031 May 1999 Revised MORs for May 1999 for Pbnps,Units 1 & 2 NPL-99-0328, Monthly Operating Repts for May 1999 for Pbnp,Units 1 & 2. with1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Pbnp,Units 1 & 2. with NPL-99-0273, Monthly Operating Repts for Apr 1999 for Point Beach Nuclear Plant,Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Point Beach Nuclear Plant,Units 1 & 2.With ML20196F3521999-04-30030 April 1999 Non-proprietary WCAP-14788, W Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wepc Point Beach Units 1 & 2 (Fuel Upgrade & Uprate to 1656 Mwt - NSSS Power) NPL-99-0193, Monthly Operating Repts for Mar 1999 for Pbnp,Units 1 & 2. with1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Pbnp,Units 1 & 2. with NPL-99-0134, Monthly Operating Repts for Feb 1999 for Pbnp,Units 1 & 2. with1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Pbnp,Units 1 & 2. with ML20207D6751999-02-22022 February 1999 Assessment of Design Info on Piping Restraints for Point Beach Nuclear Plant,Units 1 & 2.Staff Concludes That Licensee Unable to Retrieve Original Analyses That May Have Been Performed to Justify Removal of Shim Collars ML20206R9001999-01-13013 January 1999 SER Accepting Nuclear Quality Assurance Program Changes for Point Beach Nuclear Plant,Units 1 & 2 NPL-99-0008, Monthly Operating Repts for Dec 1998 for Pbnp,Units 1 & 2. with1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Pbnp,Units 1 & 2. with NPL-99-0091, 1998 Annual Results & Data Rept for Pbnps,Units 1 & 2. with1998-12-31031 December 1998 1998 Annual Results & Data Rept for Pbnps,Units 1 & 2. with ML20198C7671998-12-10010 December 1998 Safety Evaluation Accepting Licensee Proposed Alternative to ASME BPV Code,1986 Edition,Section XI Requirement IWA-2232, to Use Performance Demonstration Initiative Program During RPV Third 10-yr ISI for Plant,Unit 2 NPL-98-1006, Monthly Operating Repts for Nov 1998 for Point Beach Nuclear Plant,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Point Beach Nuclear Plant,Units 1 & 2.With ML20195J5101998-11-16016 November 1998 Proposed Revs to Section 1.3 of FSAR for Pbnp QA Program ML20198J5941998-11-0303 November 1998 1998 Graded Exercise,Conducted on 981103 NPL-98-0948, Monthly Operating Repts for Oct 1998 for Point Beach Nuclear Plant,Units 1 & 2.With1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Point Beach Nuclear Plant,Units 1 & 2.With NPL-98-0880, Special Rept:On 980913,fire Alarm Control Panels Inoperable for More That Fourteen Days.Troubleshooting of D-401 Panel Following Installation of Replacement Batteries Revealed No Apparent Cause for Spurious Alarms.Panel D-401 Restored1998-10-21021 October 1998 Special Rept:On 980913,fire Alarm Control Panels Inoperable for More That Fourteen Days.Troubleshooting of D-401 Panel Following Installation of Replacement Batteries Revealed No Apparent Cause for Spurious Alarms.Panel D-401 Restored ML20154M9121998-10-14014 October 1998 Unit 1 Refueling 24 Repair/Replacement Summary Rept for Form NIS-2 ML20154L6751998-10-14014 October 1998 Unit 1 Refueling 24 ISI Summary Rept for Form NIS-1 NPL-98-0826, Monthly Operating Repts for Sept 1998 for Point Beach Nuclear Plant,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Point Beach Nuclear Plant,Units 1 & 2.With ML20151W3851998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Pbnp Units 1 & 2 NPL-98-0653, Monthly Operating Repts for July 1998 for Point Beach Nuclear Plant,Units 1 & 21998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W4471998-07-31031 July 1998 Corrected Page to MOR for July 1998 for Pbnp Unit 2 ML20151W4541998-07-31031 July 1998 Corrected Page to MOR for July 1998 for Pbnp Unit 1 ML20236Q3161998-07-10010 July 1998 Safety Evaluation Accepting Licensee Proposed Alternative to ASME Code Requirements PTP-3-01 & PTP-3-02 ML20236L6771998-07-0707 July 1998 Safety Evaluation Approving Wepco Implementation Program to Resolve USI A-46 at Point Beach NPP Units 1 & 2 NPL-98-0558, Monthly Operating Repts for June 1998 for Pbnp,Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Pbnp,Units 1 & 2 ML20151W4261998-06-30030 June 1998 Corrected Page to MOR for June 1998 for Pbnp Unit 2 ML20151W4221998-05-31031 May 1998 Corrected Page to MOR for May 1998 for Pbnp Unit 2 NPL-98-0481, Monthly Operating Repts for May 1998 for Point Beach Nuclear Plant,Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W4011998-04-30030 April 1998 Corrected Page to MOR for April 1998 for Pbnp Unit 2 NPL-98-0356, Monthly Operating Repts for April 1998 for Point Beach Nuclear Plant,Units 1 & 21998-04-30030 April 1998 Monthly Operating Repts for April 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20217F3131998-04-17017 April 1998 Safety Evaluation Accepting Proposed Alternative to ASME Code for Surface Exam of Nonstructural Seal Welds,For Plant, Unit 1 ML20216D7071998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W3981998-03-31031 March 1998 Corrected Page to MOR for March for Pbnp Unit 2 NPL-98-0209, Special Rept Re Fire Barrier Inoperable for Greater than Seven Days.Compensatory Measures Implemented in Accordance W/Fire Protection Program Requirements During Time That Barriers Were Inoperable1998-03-30030 March 1998 Special Rept Re Fire Barrier Inoperable for Greater than Seven Days.Compensatory Measures Implemented in Accordance W/Fire Protection Program Requirements During Time That Barriers Were Inoperable ML20217A8501998-03-19019 March 1998 SER Accepting Proposed Changes Submitted on 980226 by Wiep to Pbnp Final SAR Section 1.8 Which Will Impact Commitments Made in Pbnp QA Program Description.Changes Concern Approval Authority for Procedures & Interviewing Authority ML20216J0101998-03-17017 March 1998 Safety Evaluation Accepting Third 10-yr Inservice Insp Interval Relief Request RR-1-18 for Plant NPL-98-0159, Monthly Operating Repts for Feb 1998 for Point Beach Nuclear Plant,Units 1 & 21998-02-28028 February 1998 Monthly Operating Repts for Feb 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W3891998-02-28028 February 1998 Corrected Page to MOR for Feb 1998 for Pbnp Unit 2 ML20216D7121998-02-28028 February 1998 Revised Corrected MOR for Feb 1998 for Point Beach Nuclear Plant,Unit 2 NPL-98-0084, Monthly Operating Repts for Jan 1998 for Point Beach Nuclear Plant,Units 1 & 21998-01-31031 January 1998 Monthly Operating Repts for Jan 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20198L1151998-01-0808 January 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Point Beach Nuclear Plant,Units 1 & 2 1999-09-30
[Table view] |
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+ UNITED STATES 8[ . m %,g ,
NUCLEAR REGULATORY COMMISSION t, j wassiNoTON, D. C. 20655
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET N05. 50-266 AND 50-301 REQUESTS FOR RELIEF FROM ASME CODE SECTION XI EXAMINATION REQUIREMENTS
!. BACKGROUND By letter dated October 27, 1986, Wisconsin Electric Power Company (the licensee) requested relief from the examination requirements of the 1977 Edition through Sunner 1979 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code for the shell and nozzle welds in the regenerative heat exchangers at Point Beach Nuclear Plants, Unit Nos. I and 2. The licensee provided information in support of its determination that the Code requirements were impractical to perform and proposed an alternative to the Code requirements. Inaccordancewith10CFR50.55a(g)(6)(1),the staff has evaluated the licensee's deteminations and proposed an alternative examination. The staff has concluded that the necessary findings could be made to grant relief from the Code requirements. The requests, Code requirements, the licensee's deteminations, and our bases for granting the reliefs are contained herein.
II. REQUESTS AND SUPPORTING INFORMATION A. Relief Requests RR-1-12 and RR-2-12
- to examine less than the required number of welds in the regenerative heat exchangers.
B. Exam Area
'I Class 1 Tubesheet-to-Shell Welds Class 1 Nozzle-to-Shell Welds (SeeFig.A-7 attached)
C. ASME Section XI Category & Item Number 8-B, 2.60 B-D, 3.150 D. ASME Section XI Examination Requirement A volumetric examination of 100% of all tubesheet welds and nozzle welds during the second 10-year interval, g5180353870507 p ADOCK 05000266 PDR
E. Alternative Examinations Proposed By Licensee Instead of examining all three Class 1 tubesheet-to-shell welds and all six Class 1 nozzle-to-shell welds, the licensee proposed to examine tubesheet-to-shell weld RHE-2 and nozzle-to-shell welds RHE-N1 (inlet)andRHE-N4(outlet). These welds are all located on the botton, heat exchanger (see Fig. A-7 for an outline of the RHE and weld locations). ;
F. Licensee's Reason for Limitation The regenerative heat exchanger (RHE) provides the major single source of radiation exposure accumulated during a normal refueling outage inservice inspection project. The " regenerative heat ,
exchanger" is actually three shell-and-tube heat exchangers i connected in series. The RHE is designed to recover heat from the '
reactor coolant system letdown stream by reheating the charging stream during normal operation. The letdown stream flows through the shell of the RHE and the charging stream flows through the tubes.
To ensure adequate coverage of the component welds with a minimum of exposure, the multiple stream concept should be carried out in all the welds Class 2 cap-to-shell welds. Class 2 tubesheet-to-shell welds, and two of six Class 2 norrle-to-shell welds are examined in accordance with allowances in IW8-2500 and IWC-2500 of ASME Section XI. By extending the multiple stream concept to the Class 1 '
tubesheet-to-shell and norrie-to-shell welds, a good cross-section l
of the regenerative heat exchanger would be exam < ned while a significant reduction in radiation exposure from this component would be achieved.
Following are some of the items the licensee took into consideration in the preparation of this request.
- 1. Radiation Levels Currently, the average dose rates at the regenerative heat i exchanger are:
1.5 rem /hr f 4.0 rem /hr feneral area (at 18")nsulationsurface(oncontaci 7.0res/hrshellsurface(oncontactundertheinsulation) i l
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l 1
- 2. Total Estimated Man-Rem Exposure Involved In the Examination [
Considering the tasks associated with conducting an esamination
- on a particular examination area. the following time intervals have been required in the past
0.2 man-hours for insulation removal 1 0.1 man-hours for weld cleaning and preparation
! 0.7 man-hours for conducting the examination ,
i 0.2 man-hours for insulation replacement l
' r
! Using the preceding dose rates and times, the following whole body and extremity exposures can be calculated per examination: i
- WholeBody(usinggeneralareadoserates)
I
! 1.5 res/hr for insulation removal of 0.2 man-hours = 0.3 man-rem i
- 1.5 ren/hr for weld cleaning and preparation of 0.1 ;
- man-hours = 0.15 man-rem ;
, 1.5 rem /hr for the examination of 0.7 man-hours = 1.05 man-rem j 1.5 ren/hr for insulation replacement of 0.2 man-hours = 0.3 man-rem Total Whole Body Dose Per RHE Exam = 1.8 man-res >
Extremities (hands,usingcontactdoserates): r r
- 4.0 res/hr for insulation removal of 0.2 man-hours = 0.8 man-rem 7.0 rem /hr for weld cleaning and preparation of 0.1
- man-hours
, 7.0 res/hr for the examination of 0.7 man-hours = 4.g man-rem 4.0 res/hr for insulation replacement of 0.2 man-hours = 0.8 man ren t .
i Total Whole Body Dose Per RHE Exam = 7.2 man-rom The exposure savings per inspection interval, by a reduction i
, of six examinations, would be 10.8 man-ren whole body and 43.2 l man-rem extremities.
- 3. Shielding When exposure is utilized to place lead blankets and shields
- over non-examination areas of the RHE, the general area dose i
rates are reduced by approximately 501, but the dese rates are *
' still the highest encountered during an inservice inspection project. Also, the examiner who is conducting the examination i
does not have the benefit of the shielding. .
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4
- 4. Previous Inspection P n ilts Simply stated, all 9 " cations which were recorded durian inspections to this point were found to be either insign ficant or geometric in nature. An insignificant indication is either a non-relevant indication or an indication which is equal to er greater than the examination recording level but less than the evaluation level.
- 5. Consequences of Weld Failure The consequences of a failure of one of the RHE welds have essentially been addressed in the plant's Final Safety Analysis Report (FSAR). In the FSAR, to evaluate chemical and volume control system (CVCS) safety, failures or malfunctions were assumed concurrent with a loss-of-coolant accident (LOCA) and the consequences analyzed. A LOCA and a concurrent RHE weld failure is included in the more general category of a rupture in the CVCS line inside containment. During such an occurrence, the remote-operated valve located near the main coolant loop, upstream of the RHE, is closed on low pressurizer level to prevent supplementary loss of coolant through the letdown line rupture. The RHE would also evertually be isolated, with leakage being confined to containment, in the case of
. a weld failure without a LOCA.
!!!. STAFF EVALUATION AND CONCLU$!ONS i The staff finds that the information presented by the licensee supports its conclusion that the Code-required examinations are impractical to perform on some of the welds of the regenerative heat excnangers at Point Beach Units 1 & 2. Imposition of the requirements on the licensee is not warranted when the limited volumetric examinations that can be performed on some of the upper shell welds, the levels of radiation to which the examination personnel would be subjected in performing the limited examinations, the licensee's proposed alternative examination, and the consequences of failure of a weld in the heat exchangers are considered.
Information contained in an appendix to the letter of October 27, 1986, details the extent of examination of the welds for which the licensee requests relief. These welds were previously examined to the extent practical, the extent ranging from 255 to 1005. The examination results showed insignificant or geometric indications. In addition, the consequences of failure of a regenerative heat exchanger weld were analyzed and the conclusions drawn were that the re enerative heat exchanters would be isolated from the primary loop y the remote operated valve 'n the letdown line and leakage would be conf ned to the containment.
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The licensee proposes to examine the welds in the bottom shells of the regenerative heat exchangers in accordance with Section XI requirements.
Although similar welds in the upper shells, such as RHE-6 and RHE-10 (shell-to-tubesheet welds). can be examined to Code requirements, the gain in safety would not cospensate for the additional exposure to radiation of examination personnel. The bottom shells receive reactor coolant directly from the primary loops, thereby subjecting the welds and base material to more severe operating conditions than the upper shells and making this part of the component an area most likely to develop inservice flaws. Therefore, examination of the welds in the bottom shells will provide adequate assurance of the continued structural integrity of the regenerative heat exchangers shells.
Based on the results of previous examinations perforined on the upper shells, the licensee's proposed examination of the lower shells, and the consequences of failure of a shell weld, the staff concludes that relief from the Code required examinations of the regenerative heat exchangers upper shell welds may be granted as requested.
Principal Contributor: G. Johnson Date: NAY 0 71987 l
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