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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217G2161999-10-15015 October 1999 Errata Pages 2 & 3 for Safety Evaluation Supporting Amend 168 Issued to FOL DPR-63 Issued on 990921.New Pages Change Description of Flow Control Trip Ref Cards to Be Consistent with Application for Amend ML20212D3611999-09-21021 September 1999 Safety Evaluation Supporting Amend 168 to License DPR-63 ML20210H7851999-07-29029 July 1999 Corrected SER Supporting Amend 167 to License DPR-63, Replacing Pages 4,5 & 11.New Pages Correct SE to Be Consistent with Application for Amend ML20196A3041999-06-17017 June 1999 Safety Evaluation Supporting Amend 167 to License DPR-63 ML20207G2261999-06-0707 June 1999 SER Accepting Proposed Mod to Each of Four Core Shroud Stabilizers for Implementation During Current 1999 Refueling Outage at Plant,Unit 1 ML20206U5351999-05-17017 May 1999 SER Accepting GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Plant, Units 1 & 2 ML17059C6601999-04-30030 April 1999 Safety Evaluation Accepting Licensee & Suppl ,re Alternative Repair Plan for Core Shroud Vertical Welds at Plant,Unit 1 ML17059C6111999-04-0707 April 1999 SER Accepting Util 981210 Request That NRC Approve Alternative to Performing Circumferential Shell Weld Exams on RPV Welds at Nine Mile Point Nuclear Station,Unit 1 ML17059C6031999-04-0202 April 1999 Safety Evaluation Supporting Alternative GVRR-6 Re IST of Safety & Relief Valves in Steam or Compressible Fluid Svc ML17059C5951999-03-18018 March 1999 Safety Evaluation Supporting Amend 86 to License NPF-69 ML20204C8001999-03-16016 March 1999 Safety Evaluation Supporting Amend 165 to License DPR-63 ML17059C5701999-02-18018 February 1999 Safety Evaluation Supporting NMPC Responses to NRC Bulletin 95-002, Unexpected Clogging of Rhrp Strainer While Operating in Suppression Pool Cooling Mode ML17059C5611999-02-17017 February 1999 Safety Evaluation Approving Inservice Testing Program Relief Request GVRR-01 ML17059C4331998-12-14014 December 1998 Safety Evaluation Concluding That Proposed Restructuring of Rg&E by Creation of Holding Company Will Not Adversely Affect Financial Qualifications of Rg&E Re Operation & Decommissioning of Nine Mile Point Unit 2 ML17059C4231998-12-11011 December 1998 SER Approving Restructuring of NMP Corp by Creation of Holding Company ML17059C3941998-12-0303 December 1998 Safety Evaluation Supporting Amend 84 to License NPF-69 ML17059C3821998-11-25025 November 1998 Safety Evaluation Supporting Amend 164 to License DPR-63 ML20155E2001998-11-0202 November 1998 Safety Evaluation Approving NMP 980227 Request for Extension of Reinspection Interval for Core Shroud Vertical Welds at NMP1 from 10,600 Hours to 14,500 Hours of Hot Operation ML17059C2871998-10-15015 October 1998 Safety Evaluation Accepting Results of Ultrasonic Insp of Core Shroud Welds During 1998 Refueling Outage at Nmpns,Unit 2 & Provided Flaw Evaluation for Continued Operation of Unit 2 for One Operating Cycle ML20154D8401998-10-0505 October 1998 Safety Evaluation Accepting Proposed Changes Related to PT Limits in Plant,Unit 1 TSs ML17059C2511998-09-18018 September 1998 Safety Evaluation Supporting Amend 163 to License DPR-63 ML17059C2141998-08-27027 August 1998 Safety Evaluation Accepting Topical Rept GENE-A13-00360-02 & as Ref Source for NMPNS Unit 1,for Purpose of Requesting TS Change to Implement Option II & Implementing Associated Hardware Changes NUREG-1407, Safety Evaluation on IPEEE for Facility1998-08-12012 August 1998 Safety Evaluation on IPEEE for Facility ML17059C1611998-07-23023 July 1998 Safety Evaluation Accepting Licensee Use of Alternatives to ASME Code XI Exam Requirements for Sixth Refueling Outage, Unit 2 ML17059C1501998-07-19019 July 1998 Safety Evaluation Accepting Proposed Order to Indirect Transfer of Facility Operating License NPF-69 to Extent Held by Util,To Unnamed Holding Company to Be Created by Util IAW Electric Industry Restructuring Goals ML17059C1061998-06-25025 June 1998 Safety Evaluation Accepting Flaw Indication on Feedwater nozzle-to-safe End Weld Butter Indication for Plant,Unit 2 ML17059C0741998-06-0404 June 1998 Safety Evaluation Supporting Amend 82 to License NPF-69 ML17059C0201998-05-23023 May 1998 Safety Evaluation Supporting Amend 161 to License DPR-63 ML17059C0041998-05-20020 May 1998 SER Accepting Util 971215 Appication for Amend to License NPF-69,revising Critical Power Ratio Safety Limit in Section 2.1.2 of TS from 1.07 to 1.09 for Two Recirculation Loop Operation ML17059B9711998-04-16016 April 1998 Safety Evaluation Supporting Amend 81 to License NPF-69 ML17059B9721998-04-15015 April 1998 Errata to Safety Evaluation Supporting Amend 80 to License NPF-69.Replacement Pages 2,3,7 & 8 Incorporate Editorial Corrections & Are Consistent with Application for Amend ML17059B9471998-03-31031 March 1998 Safety Evaluation Supporting Amend 80 to License NPF-69 ML20217F4341998-03-19019 March 1998 SER Related to Proposed Restructuring New York State Electric & Gas Corp,Nine Mile Point Nuclear Station,Unit 2 ML17059B9331998-03-16016 March 1998 Safety Evaluation Supporting Amend 79 to License NPF-69 ML17059B9011998-02-19019 February 1998 Safety Evaluation Supporting Amend 160 to License DPR-63 ML20198H9941997-12-29029 December 1997 SE Supporting Approval of Application Re Long Island Power Authority Aquisition of Long Island Lighting Co,Subject to Discussed Condition ML17059B5421997-05-0808 May 1997 Safety Evaluation Supporting Results of Reinsp of Util Core Shroud for Plant,Unit 1 ML17059B5271997-04-30030 April 1997 Safety Evaluation Accepting Flaw Evaluation of Recirculation Line Weld 32-WD-050 for Plant,Unit 1 ML17059B4561997-03-0505 March 1997 Safety Evaluation Accepting Relief Request for Valve Inservice Testing Program ML17059B4491997-03-0303 March 1997 Safety Evaluation Supporting Reinspection of Core Shroud During RFO-14 Re License DPR-63 ML17059B4461997-03-0303 March 1997 Safety Evaluation Accepting Proposed Mods to Shroud Repair Assembly ML17059B4291997-02-10010 February 1997 Safety Evaluation Supporting Amend 159 to License DPR-63 ML20197C4771997-01-0303 January 1997 Safety Evaluation Supporting Nine Mile Point Unit 1 Reactor & Turbine Building Blowout Panels ML20197C4541997-01-0202 January 1997 Safety Evaluation Supporting Resolution of Nine Mile Point Reactor/Turbine Building Pressure Relief Panel Outside Design Basis Issue ML17059B3641996-12-12012 December 1996 Safety Evaluation Supporting Amend 78 to License NPF-69 ML17059B3691996-12-12012 December 1996 Safety Evaluation Supporting Amend 158 to License DPR-63 ML17059B2891996-09-17017 September 1996 Safety Evaluation Supporting Amend 77 to License NPF-69 ML17059B2591996-08-29029 August 1996 Corrected Page 4 of SE Supporting Amend 74 to License NPF-69 ML17059B2581996-08-28028 August 1996 Safety Evaluation Supporting Amend 76 to License NPF-69 ML17059B2631996-08-27027 August 1996 Safety Evaluation Supporting Amend 75 to License NPF-69 1999-09-21
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G2161999-10-15015 October 1999 Errata Pages 2 & 3 for Safety Evaluation Supporting Amend 168 Issued to FOL DPR-63 Issued on 990921.New Pages Change Description of Flow Control Trip Ref Cards to Be Consistent with Application for Amend ML20216J9251999-09-30030 September 1999 Suppl to Special Rept:On 990621,11 Containment Hydrogen Monitoring Sys Chart Recorder Was Indicating Below Normal Operating Range.Caused by Excessive Wear on Valve Body & Discs of Bypass Pump.Sample Pump Replaced ML20217K4631999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Nine Mile Point, Unit 1.With ML20212F7301999-09-21021 September 1999 Special Rept:On 990907,CR Operators Declared 12 Containment Hydrogen Monitoring Sys Inoperable for Planned Maint.Cause of Low Flow Condition Was Determined to Be Foreign Matl. Replaced Sample Pump Valve Discs ML20212D3611999-09-21021 September 1999 Safety Evaluation Supporting Amend 168 to License DPR-63 ML20212B9081999-09-14014 September 1999 Special Rept:On 990901, 12 Containment Hydrogen Monitoring Sys Was Declared Inoperable for Planned Maint.Caused by Planned Maint Being Performed as Corrective Action.Check Valves with O Rings Were Replaced ML20212C4601999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Nine Mile Point Nuclear Station,Unit 1.With ML20216F5141999-08-31031 August 1999 Rept on Status of Public Petitions Under 10CFR2.206 05000220/LER-1999-005-01, :on 990801,RT During Plant Startup on IRM Spiking Was Noted.Caused by electro-magnetic Interference. Inspected,Cleaned,Burnished & Tested IRM 11 Range Switch. with1999-08-30030 August 1999
- on 990801,RT During Plant Startup on IRM Spiking Was Noted.Caused by electro-magnetic Interference. Inspected,Cleaned,Burnished & Tested IRM 11 Range Switch. with
05000220/LER-1999-004-01, :on 990723,reactor Scram Was Noted.Caused by Mechanical Pressure Regulator Suppressor Valve Failure. Cleaned Pressure Suppressor Valve & Provided Guidance for Valve Disassembly & Insp.With1999-08-23023 August 1999
- on 990723,reactor Scram Was Noted.Caused by Mechanical Pressure Regulator Suppressor Valve Failure. Cleaned Pressure Suppressor Valve & Provided Guidance for Valve Disassembly & Insp.With
ML20210U4591999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Nine Mile Point, Unit 1.With ML20210H7851999-07-29029 July 1999 Corrected SER Supporting Amend 167 to License DPR-63, Replacing Pages 4,5 & 11.New Pages Correct SE to Be Consistent with Application for Amend ML20209D0291999-07-0202 July 1999 Special Rept:On 990621,operator Identified That Number 11 Hydrogen Monitoring Sys (Hms) Chart Recorder Was Indicating Below Normal Operating Range.Cause Indeterminate.Licensee Will Complete Troubleshooting of Subject Hms by 990709 ML20210B9081999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Nine Mile Point Unit 1.With ML20196A3041999-06-17017 June 1999 Safety Evaluation Supporting Amend 167 to License DPR-63 ML20209F8811999-06-0808 June 1999 Rev 1 to NMP Unit 1 COLR for Cycle 14 ML20207G2261999-06-0707 June 1999 SER Accepting Proposed Mod to Each of Four Core Shroud Stabilizers for Implementation During Current 1999 Refueling Outage at Plant,Unit 1 ML20196E2111999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Nmp,Unit 1.With ML20207B0241999-05-18018 May 1999 Safety Evaluation of Topical Rept TR-107285, BWR Vessel & Intervals Project,Bwr Top Guide Insp & Flaw Evaluation Guidelines (BWRVIP-26), Dtd December 1996.Rept Acceptable ML20206U5351999-05-17017 May 1999 SER Accepting GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Plant, Units 1 & 2 ML20196L2301999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Nmp,Unit 1.With ML17059C6601999-04-30030 April 1999 Safety Evaluation Accepting Licensee & Suppl ,re Alternative Repair Plan for Core Shroud Vertical Welds at Plant,Unit 1 05000220/LER-1999-003-01, :on 990320,noted That Plant Operated with Average Planear LHGR Exceeding TS Limits.Caused by Inadequate Interface Design Between Color Graphics Terminal & 3-D Monicore Sys.Lowered Reactor Power.With1999-04-22022 April 1999
- on 990320,noted That Plant Operated with Average Planear LHGR Exceeding TS Limits.Caused by Inadequate Interface Design Between Color Graphics Terminal & 3-D Monicore Sys.Lowered Reactor Power.With
ML17059C6111999-04-0707 April 1999 SER Accepting Util 981210 Request That NRC Approve Alternative to Performing Circumferential Shell Weld Exams on RPV Welds at Nine Mile Point Nuclear Station,Unit 1 ML17059C6031999-04-0202 April 1999 Safety Evaluation Supporting Alternative GVRR-6 Re IST of Safety & Relief Valves in Steam or Compressible Fluid Svc ML20205L0541999-04-0101 April 1999 Nonproprietary Replacement Pages to HI-91738,consisting of Section 5.0, Thermal-Hydraulic Analysis ML20205S5701999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for NMP Unit 1.With 05000220/LER-1999-002, :on 990217,noted That ASME Code Preservice Exams Had Not Been Performed on Emergency Condensers.Caused by Misinterpretation of ASME Code Requirements.Required Preservice Exams Were Completed on 990218.With1999-03-19019 March 1999
- on 990217,noted That ASME Code Preservice Exams Had Not Been Performed on Emergency Condensers.Caused by Misinterpretation of ASME Code Requirements.Required Preservice Exams Were Completed on 990218.With
ML17059C5951999-03-18018 March 1999 Safety Evaluation Supporting Amend 86 to License NPF-69 ML20204C8001999-03-16016 March 1999 Safety Evaluation Supporting Amend 165 to License DPR-63 05000410/LER-1999-001-01, :on 990212,NMP2 Was Outside Design Basis Due to Safe SD SW Pump Bay Unit Coolers Being Oos.Caused by Inadequate Managerial Methods.Interim ACs Were Set for Safe SD Equipment.With1999-03-15015 March 1999
- on 990212,NMP2 Was Outside Design Basis Due to Safe SD SW Pump Bay Unit Coolers Being Oos.Caused by Inadequate Managerial Methods.Interim ACs Were Set for Safe SD Equipment.With
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 ML20207G2671999-03-0101 March 1999 Special Rept:On 990315,Nine Mile Point,Unit 1 Declared Number 12 Containment Hydrogen Monitoring Sys Inoperable. Caused by Degraded Encapsulated Reed Switch within Flow Switch FS-201.2-1495.Technicians Replaced Flow Switch ML20204C9971999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Nine Mile Point,Unit 1.With ML20207E9311999-02-26026 February 1999 Part 21 Rept Re Sprague Model TE1302 Aluminum Electrolytic Capacitors with Date Code of 9322H.Caused by Aluminum Electrolytic Capacitors.Affected Capacitors Replaced 05000220/LER-1999-001, :on 990125,noted That Plant Operated Outside Design Basis Due to Failure to Revise Satellite pre-fire Plans.Satellite Copies of pre-fire Plans Were Revised & Procedure Re pre-fire Plans Was Revised.With1999-02-24024 February 1999
- on 990125,noted That Plant Operated Outside Design Basis Due to Failure to Revise Satellite pre-fire Plans.Satellite Copies of pre-fire Plans Were Revised & Procedure Re pre-fire Plans Was Revised.With
ML17059C5701999-02-18018 February 1999 Safety Evaluation Supporting NMPC Responses to NRC Bulletin 95-002, Unexpected Clogging of Rhrp Strainer While Operating in Suppression Pool Cooling Mode ML17059C5611999-02-17017 February 1999 Safety Evaluation Approving Inservice Testing Program Relief Request GVRR-01 ML17059C5501999-01-31031 January 1999 Rev 0 to MPR-1966(NP), NMP Unit 1 Core Shroud Vertical Weld Repair Design Rept ML20199M0891999-01-22022 January 1999 Part 21 Rept Re Failure of Square Root Converters.Caused by Failed Aluminum Electrolytic Capacitory Spargue Electric Co (Model Number TE1302 with Mfg Date Code 9322H).Sent Square Root Converters Back to Mfg,Barker Microfarads,Inc ML20206P2391998-12-31031 December 1998 Special Rept:On 981222,operators Removed non-TS Channel 12 Drywell Pressure Recorder & Associated TS Pressure Indicator from Svc.Caused by Intermittent Measuring Cable Connection in non-TS Recorder Circuitry.Replaced Cable ML20210R8441998-12-31031 December 1998 1998 Annual Rept for Energy East ML20199K9331998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Nine Mile Point Nuclear Station,Unit 1.With ML20206P2421998-12-30030 December 1998 Special Rept:On 981219,number 12 Hydrogen Monitoring Sys (Hms) Was Declared Inoperable When Operators Closed Valve 201.2-601.Caused by Indeterminate Failure of Valve 201.2-71. Supplemental Rept Will Be Submitted After Valve Is Repaired ML20198M3571998-12-23023 December 1998 Special Rept:On 981210,operators Declared Number 11 Inoperable,Due to Failure of CR Chart Recorder.Caused by Inverter Board in Power Supply Circuitry of Recorder Due to Component Aging.Maint Personnel Replaced Failed Inverter ML17059C4331998-12-14014 December 1998 Safety Evaluation Concluding That Proposed Restructuring of Rg&E by Creation of Holding Company Will Not Adversely Affect Financial Qualifications of Rg&E Re Operation & Decommissioning of Nine Mile Point Unit 2 ML17059C4231998-12-11011 December 1998 SER Approving Restructuring of NMP Corp by Creation of Holding Company 05000220/LER-1998-019-01, :on 981110,ASME Section XI ISI Was Missed. Caused by Cognitive Error Involving Technical Inaccuracies. Util Will Revise Second 10-yr Interval ISI Program Plan to Incorporate Visual Exam Requirements.With1998-12-10010 December 1998
- on 981110,ASME Section XI ISI Was Missed. Caused by Cognitive Error Involving Technical Inaccuracies. Util Will Revise Second 10-yr Interval ISI Program Plan to Incorporate Visual Exam Requirements.With
ML17059C3941998-12-0303 December 1998 Safety Evaluation Supporting Amend 84 to License NPF-69 ML20198D9361998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Nine Mile Point,Unit 1.With 1999-09-30
[Table view] |
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UNITED STATES p
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,j NUCLEAR REGULATORY COMMISSION O..... /g WASHINGTON, D.C. 30666 4 001 p
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION MODIFICATION OF CORE SHROUD TIE ROD UPPER SPRING ASSEMBLIES j
i NINE MILE POINT NUCLEAR STATION. UNIT NO.1 f
DOCKET NO. 50-220
1.0 INTRODUCTION
By letter dated May 21,1999, as supplemented by letter dated May 28,1999, Niagara Mohawk Power Corporation (NMPC or licensee) proposed a modification to each of the four core shroud stabilizers (a.k.a., tie rod assemblies) for implementation during the current 1999 refueling outage at Nine Mile Point Nuclear Station, Unit 1 (NMP1). NMPC provided additional information to the U.S. Nuclear Regulatory Commission (NRC) staff during a telephone call on l
May 27,1999 (see summary of telephone conversation dated June 3,1999).
l Pursuant to 10 CFR 50.55a(g)(4), the components must meet the requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).
Because the proposed modification is not included under the definition of repair or replacement specified in Section XI of the ASME Code, the proposed modification was submitted to the NRC l
staff for review and approval as an alternative repair pursuant to Title 10 of the Code of Federal Reaulations (10 CFR) Section 50.55a(a)(3)(i).
2.0 BACKGROUND
The proposed modification results from NMPC's visual examination of the tie rod assemblies during the current refueling outage (RFO15). The examination revealed that a 3/8-inch cap screw connecting the spring bracket to the upper spring in one of the tie rod assemblies had l
failed and that the broken portion had dislodged from the upper spring assembly. In addition, NMPC's visual examination of the reactor pressure vessel (RPV) cladding revealed areas of scratches and some evidence of wear at the locations where the upper spring of each of the four tie rod assemblies contacts the RPV cladding. NMPC determined that the scratches and extent of wear were acceptable.
The root cause of the cap screw failure was intergranular stress-corrosion cracking (IGSCC) in the alloy X-750 material driven by large, sustained differential thermal expansion stress due to fastening of dissimilar materials with the cap screw. A potential contributing cause was the sustained stresses that were attributed to the torquing of the cap screw associated with the Enclosure
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9906110066 990607 PDR ADOCK 05000220 l
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.. original assembly of the tie rods. A second potential contributing cause was the stresses associated with friction between the vessel wall and the upper spring contact points.
The proposed modification would replace the design function of the failed cap screw and other
. cap screws that have the potential for future failure. The modification includes changes to each end of the upper spring for each of the four tie rods and contains provisions to prevent the creation of loose parts in the RPV due to the potential for failure of the subject cap screws. In addition, the modification includes rounding off the leading edges of the upper wedge and l
upper contact of the upper spring for the four tie rods as a preventative measure to reduce stresses on the tie rod assemblies and to reduce the likelihood of future wear on the cladding of the RPV.
The proposed modification to the upper spring, including the tie rod assemblies, are illustrated in Figures 1,2,3, and 4 attached to this safety evaluation (SE). The upper spring modification, shown in Figures 2,3, and 4, involves the addition of a clamp fabricated from 300 series stainless steel material, secured to attach the existing upper spring bracket to the upper spring by two bolts, consisting of XM-19 material, and installed perpendicular to the plane of maximum l
thermal expansion. Currently, four 3/8" cap screws secure the upper spring bracket to the upper spring. The two XM-19 bolts replace the function of the four existing 3/8" cap crews.
1 The XM-19 threaded bolts are staked in place by type 316/316L locking pins. The clamp also functions to prevent the existing 3/8" cap screws from loosening or dislodging and becoming loose parts.
)
3.0 EVALUATION 3.1 Structural Evaluation NMPC stated that a structural evaluation of the proposed modification was conducted. NMPC's analysis and conclusions were reviewed by the NRC staff, and are summarized in the following three paragraphs.
The significant loads on the bolts of the upper tie rod spring assembly consist of loads due to differential thermal expansion loads caused by fastening of dissimilar materials, loads due to installation torque, and loads due to friction or potential binding between the RPV and the contact wedges.
Postulated dynamic events during accident conditions do not impose any additional loads.
However, for the purpose of checking the bearing at the contact wedges, seismic loads were included along with the upset thermal condition loads. The detailed analysis for the bolts was performed for the normal and upset operating conditions. The analysis of the bolts was based upon the use of X-750 and XM 19 materials for the bolt. Although both materials were found acceptable, XM-19 was selected. In the evaluation, it is assumed that the currently remaining j
bolts (X-750) are ineffective. The calculations were performed using strength of material equations, thermal and mechanical properties, and allowable values in accordance with the ASME Code, Section lil, Subsection NG.
The tensile stress in ths bolt was determined to be 13.7 ksi (the larger of the normal and thermal upset conditions). This is well within the allowable stress of 35.4 ksi which is 0.9 Sy at i
l
the upset temperature for the XM-19 material bolt. The allowable stresses for the XM-19 material were based upon annealing at a temperature of 2025-2075 *F. Bolt head shear stress was evaluated for the upper contact bolts and shown to be acceptable. Fatigue was also assessed in the analysis. The allowable fatigue cycles were determined to be 700,000 cycles.
The actual number of thermal cycles (startup and shutdown) was conservatively assumed as 200, which is much smaller than the allowable 700,000 cycles. Thus, fatigue stresses were determined to be insignificant. The bolts are threaded into the nut plate. The only significant l
loading on the nut plate is due to the tensile load in the bolt. The thread shear in the nut plate due to the bolt load was determined to be less than 3.5 ksi, which is well within the allowable value of 11.3 ksi, which is 0.6 Sy at the thermal upset temperature for the plate material.
l The NRC staff has reviewed NMPC's structural evaluation and finds it to be acceptable. On the basis of this review, the NRC staff concludes that the design of the repair modification satisfies the structural requirements of the ASME Code and, therefore, provides an acceptable level of quality and safety.
3.2 Materials Evaluation i
in its submittal to the NRC staff, NMPC states that the clamps will be fabricated from type 316L stainless steel materials and the bolts from XM-19 materials. NMPC indicated that the XM-19 materials for the bolts were selected because of material availability and because its thermal expansion coefficient is similar to that of the type 316L stainless steel parts to which it is adjoined, thereby minimizing the thermal expansion tensile stresses. The locking pins used to prevent the loosening of pre-torqued XM-19 bolts will be made from type 316/316L stainless steel materials. All the components will be fabricated from materials in a solution-annealed condition and the materials will be tested to ensure that they were not sensitized. To minimize susceptibility to IGSCC due to surface cold work, requirements for cold work control in machining the threads are included in the fabrication specification. To eliminate potential stress corrosion cracking due to surface contamination, chemical controis are imposed during the process of fabrication to ensure that there is no surface contamination such as by heavy metals and chlorine or sulfur compounds.
The NRC staff has determined that the materials (type 316/316L and XM-19) selected for the repair of components are ecceptable because they are ASME Code-approved materials. The i
solution annealed condition of these materials, together with the machining and chemical controls imposed during fabrication, will provide the repair of components with adequate l
resistance to IGSCC. Various BWR internals and other repair components made of these materials have demonstrated successful service experience in the BWR environment.
3.3 Loose Parts Considerations 4
In its submittal, NMPC states that the repair uses locking methods for threaded connections.
l _
All parts will be captured and held in place by a method that will last for the design life of the repair. The modified upper springs are designed to minimize the potential for loose parts inside the RPV, and the addition of the clamps are intended to prevent the existing 3/8" cap screws from becoming loose parts. The threaded bolts are captured and held in place by interference-fit locking pins that stake the bolts. The locking pins are prevented from backing out of their installed holes by reducing the diameter of the hole edge after installation of the pins.
.7 4-Therefore, the NRC staff finds that the proposed modification includes acceptable provisions to preclude the occurrence of loose parts. -
4.0 CONCLUSION
On the basis of its review of the analysis discussed above, the NRC staff finds that the structural design of the repair modification satisfies the structural requirements of the ASME Code, Section lil, Subsection NG, " Core Support Structures," and is, therefore, acceptable.
The NRC staff finds that the materials selected for component repair are specified as approved materials by the ASME Code and are, therefore, acceptable. Acceptable provisions are included to preclude the occurrence of loose parts. Accordingly, the NRC staff concludes that the proposed modification provides an acceptable level of quality and safety. Accordingly, it is 1
authorized in accordance 10 CFR 50.55a(a)(3)(i).
]
b Attachments:
1
- 1. Figure 1, Core Shroud Stabilizers
- 2. Figure 2, Upper Spring
- 3. Figure 3, Upper Clamp installation
- 4. Figure 4, Lower Clamp Assempty PrincipalContributors: D. Hood
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J. Rajan W. Koo Date: June 7,1999
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