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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
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NUCLEAR REGULATORY COMMISSION g WASHINGTON, D.C. 20066 4001 4
9 . . . . . ,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REVIEW OF PROPOSED ALTERNATIVE TO THE ASME CODE FOR SURFACE EXAMIN'ATION OF NONSTRUCTURAL SEAL WELDS WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT. UNIT 1 DOCKET NO. 50-266
1.0 INTRODUCTION
Title 10, Code of Federal Regulations (CFR), 50.55a(c)(4) states that for a nuclear power plant whose construction permit was issued prior to May 14,1984, the applicable Code edition and Addenda for a component of the reactor coolant pressure boundary continue to be that Code
! Edition and Addenda that were required by Commission regulations for such components at the time of issuance of the construction permit.
10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (c) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
By letter dated April 3,1998, Wisconsin Electric (WE or the licensee) Power Company stated its intent to remove the part-length control rod drive mechanisms (CRDMs), cut the reactor vessel penetrations for the part-length CRDMs at locations E7, G5, G9, and 17, install threaded caps, and seal weld the caps. In accordance with 10 CFR 50.55a(c)(4), the modification would be to the requirements of the original 1965 Edition of the Code for the reactor coolant pressure boundary. WE has proposed an attemative to the surface examination requirements of paragraph N-518.4 of the 1965 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. The proposed attemative would be used in conjunction with the four nonstructural seal welds on the threaded caps for the Unit i reactor vessel head penetrations.
The welds will be used to ensure leak tightness of the thread joints holding the new caps Because the threads constitute the pressure boundary, the seal weld is nonstructural.
The staff has reviewed and evaluated the licensee's request and supporting information on the proposed attemative to the Code requirements for Point Beach Unit 2, pursuant to the provisions of 10 CFR 50.55a(a)(3)(i).
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2.0 DISCUSSION
Component Identification:
Reactor vessel penetrations E7, G5, G9, and 17, for the part-length CRDM housings.
l Code Requirement.
l ASME Section 111,1965 Edition, paragraph N-518.4, specifies a surface examination of welded l attachments. Paragraph N-518.4 of the 1965 ASME Code, Section lil, requires that attachments (welds) to the pressure boundary be inspected by means of a liquid penetrant examination (PT). For PT examinations, the acceptance criteria of paragraph N-627 must be met. The most stringent among the acceptance criteria is the requirement for "no linear indications." Later editions of the Code define 1/16 inch as the smallest length of a relevant l linear indication.
l Alternate Reauirement: (As stated) l In accordance with the provisions of 10 CFR Part 50, Section 50.55a(a)(3), Wisconsin Electric l (WE) Power Company proposes the following:
- 1. Use of a controlled automatic welding process.
- 2. Observation of the weld puddle / deposit via a 8X camera during the welding process.
- 3. A final visual examination of the weld surface using the same 8X camera.
- 4. Performance of a VT-2 inspection of the seal weld and CRDM penetration cap for leakage during the reactor coolant system (RCS) leak test. The RCS leak test is performed at 2085 psig,100 psig above the normal operating pressure of 1985 psig.
- 5. Authorized Nuclear Inservice inspector approval of alternative testing a.1d NIS-2 acceptance.
l Reason for Proposed Alternate Reauirement:
WE stated that the PT examination of the welds would be difficult. The seal welds will be in a high radiation area of approximately 1000 millirem / hour. Additionally, access to the weld locations is difficult due to the limited clearance between adjacent CRDMs. The separation between the housings is approximately 7 inches. This is not adequate clearance to gain complete access to perform the PT examination. Surface preparation (grinding) of the welds, PT examination, and subsequent cleanup would be difficult to execute properly, would be time consuming, and would incur substantial personnel radiation exposure.
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. s 3.0 - EVALUATION Section lli of the ASME Code,1965 Edition, specifies a surface examination of weld-repaired areas (paragraph N-514.2) or welded attachments (paragraph N-518.4). For PT examinations, the acceptance criteria of paragraph N-627 must be met. The most stringent among the acceptance criteria is the requirement for "no linear indications." Later editions of the Code define 1/16 inch as the smallest length of a relevant linear indication.
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! The proposed alternative would include an 8-power visual examination (by means of a remote l video camera) of the in-process automatic welding and a post-weld visual examination with the same camera. Although the seal weld is nonstructural, the licensee had an analysis performed that treated the weld as a pressure boundary in order to calculate a critical flaw size (critical l through-wall axial flaw length of 4.17 inches and critical through-wall circumferential flaw length l of 6.95 inches). The results were used as a reference for the evaluation of the effectiveness of the visual examination with the camera. The results of the analysis demonstrated that the ,
predicted critical flaw size ir, all cases is of significant length and therefore the proposed -l 8-power visual examination will be able to reliably detect much smaller flaws than the calculated .
critical flaw size. A VT-2 visual examination of the welds would also be performed during a l hydrostatic test of the reactor This inspection assures that the function of the seal weld is i being satisfied. !
I The licensee subminea a test report giving the results of a resolution test of the camera equipment used by the welding contractor during the weld repairs. In the test, a wire .
0.0005-inch in diameter and another 0.001-inch in diameter, both 0.4-inch long, were taped to I the surface adjacent to a mockup of the production welds. The wire was filmed using the weld ,
head lighting for illumination. Review of the tape demonstrated the visibility of the test wires :
with the camera system.
As a further process control during welding, the same video camera will be employed to monitor the weld puddle during performance of the production welds; This technique is now commonly .
employed in the industry with positive results. It enables the welding operator to further verify the welding process, take corrective actions during the course of welding, and to identify potential problem locations prior to weld completion and performance of weld acceptance examinations. With the additional process monitoring this method provides, the probability of undetected weld defects is substantially diminished.
The use of a controlled automatic welding process, observation of the weld puddle / deposit during the welding process, a final visual examination of the weld surface using the camera, l and a VT-2 examination of the weld and the cap for leakage during the post-outage hydrostatic test will provide assurance of the integrity of the seal welds. Therefore, the proposed alternative will provide an acceptable level of quality and safety.
4.0 CONCLUSION
Under the provisions of 10 CFR 50.55a(a)(3)(i), the staff has determined that the licensee's proposed attemative to the ASME Code for surface examination (PT) of the seal welds on threaded caps for the Unit i reactor vessel head penetrations for the part-length CRDMs will l
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4 l provide an acceptable level of quality and safety. The proposed attemative inspection is l authorized for the four stated repairs at Point Beach Nuclear Plant Unit 1. Authorization of the l proposed attemative does not extend to future use without prior staff review. If similar repairs l are necessary for the Unit 2 penetrations in the future, a separate relief request must be submitted.
i Principal Contributor: A. Kugler Date: April 17, 1998 l
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