NRC Generic Letter 1994-03: Difference between revisions
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==Purpose== | ==Purpose== | ||
The purpose of this Generic Letter is to request that each addressee: (1)inspect the core shrouds in their BWR plants no later than the next scheduledrefueling outage, and perform an appropriate evaluation and/or repair based onthe results of the inspection; and (2) perform a safety analysis supportingcontinued operation of the facility until inspections are conducted.BackgroundIntergranular stress corrosion cracking (IGSCC) of BWR internal components hasbeen identified as a technical issue of concern by both the NRC staff and theindustry. The core shroud is among the list of internals susceptible toIGSCC. Identification of cracking at the circumferential beltline regionwelds in several plants during 1993 led to the publication of NRC InformationNotice (IN) 93-79, issued on September 30, 1993. Several licensees haverecently inspected their core shrouds during Spring 1994 planned outages andhave identified extensive cracking at the circumferential welds. Theseinspection findings are causing the NRC staff and industry to re-evaluate thesignificance of this issue. Due to the 3600 degree extent of the cracking,and the location at a lower elevation where extensive cracking had not beenpreviously observed (e.g., H5 in the attached figures), the inspections andanalyses performed for Dresden Unit 3 and Quad Cities Unit 1 (Ref. 1, 2) areespecially noteworthy. NRC has issued IN 94-42 on June 7, 1994, andSupplement 1 to IN 94-42 on July 19, 1994, on cracking in the lower region ofthe core shroud found at Dresden Unit 3 and Quad Cities Unit 1. In additionto the core shroud, NRC has an overall concern with cracking of BWR internalsand encourages licensees to work closely with the BWR Owners Group (BWROG) oncoordination of inspections, evaluations and repair options for internalscracking.POP, ocK 050000o 3 P (AOat9407210200 ot 0 d Lo1 4 -i/IC e / ) c Le7J GL 94-03July 25, 1994 DiscussionEVALUATION OF RECENT INSPECTION EXPERIENCEBy letter dated April 5, 1994 (Ref. 3), the BWROG submitted to NRC genericguidance on the evaluation of BWR core shrouds. This guidance included aninspection strategy that was based on examination of the results of plantinspections up to that time. This inspection strategy was founded on IGSCCsusceptibility rankings and involved focusing the examinations on the uppershroud welds (e.g., H2, H3). Enhanced visual (VT-1) or ultrasonic (UT)methods of portions of the upper shroud welds were recommended for the initialexaminations. However, in light of the recent inspections at Dresden 3 andQuad Cities 1, the BWROG is re-evaluating the applicability of the inspectionguidance. Cracking extending 3600 around the shroud circumference wasobserved at the H5 weld location at both Dresden 3 and Quad Cities 1.Extensive cracking at this location had not been observed previously and wouldnot have been expected based on the BWROG guidance. As all BWR plants havenot performed inspections of their core shrouds, and since the core shroudcracking phenomenon is dependent on operating time and plant specificconditions, additional inspections are necessary to verify that conditionspotentially worse than those already identified do not exist at other plantsand that appropriate corrective actions are taken.SIGNIFICANT PARAMETERS AFFECTING CORE SHROUD CRACKINGThe BWROG has also previously discussed the significant parameters known toaffect the susceptibility of core shrouds to IGSCC (Ref. 3). These parametersinclude, but are not necessarily limited to materials, fabrication andresidual stresses, water chemistry, and fluence. Within and among these broadcategorizations, there exists sufficient variability to make an accurateprediction of IGSCC susceptibility difficult on a generic basis. While theNRC recognizes the usefulness of these categorizations, susceptibility tocracking, or lack thereof, needs to be demonstrated on a plant-specific basis.SIGNIFICANCE OF PART THROUGH-WALL 3600 CRACKINGNRC has assessed the safety significance of part through-wall 3600 core shroudcracking and has concluded that, for the most significant cracks found to date(up to 3600 circumferential extent), the structural margins required by theASME Boiler and Pressure Vessel Code pursuant to Section 50.55a of Title 10 ofthe Code of Federal Regulations [10 CFR 50.55a(g)] were maintained, therebyproviding assurance that the shrouds would have remained intact even underpostulated accident conditions. The ASME Code, Section XI, SubarticleIWB-2500, categories B-N-1 and B-N-2 specify examination and acceptancerequirements for reactor internals and core support structures, including thecore shroud. Paragraph IWB-3520 is referenced as the acceptance standard forintegrally welded core support structures and reactor interior attachments.By letter dated July 13, 1994 (Ref. 4), the BWROG submitted a response toprevious NRC staff questions regarding the susceptibility of BWRs to safety-significant shroud cracks. In this response, the BWROG provided an evaluation GL 94-03July 25, 1994 of the cracking that has been observed in plants which have inspected theirshrouds. The plants which have experienced the most extensive cracking haveoperated for longer than 8 years and had moderate to high coolant conductivityover the first 5 cycles of operation. The BWROG evaluation indicates that thestructural margins for plants most susceptible to cracking would be maintainedfor at least one more cycle of additional operation at current conductivitylevels. However, the BWROG notes that the uncertainties in the assumptionslead to the conclusion that while development of cracks that would not satisfythe ASME Code factors of safety is unlikely, such an occurrence cannot beruled out. Part of the purpose of this generic letter is to ascertain thelikelihood of such an occurrence for each BWR plant and to take appropriatecorrective action(s).SAFETY SIGNIFICANCE OF POSTULATED ACCIDENTS WITH 3600 THROUGH-WALL CRACKSIn order to assess the significance of potential cracking worse than thatobserved to date, NRC has evaluated the safety implications of a postulated3600 circumferential separation of the shroud for which the ASME Code safetymargins are clearly not met. Based on this evaluation, NRC has determinedthat 360° through-wall cracking of the core shroud may not be identified undernormal operating conditions, depending on the elevation of the cracking in theshroud. At the upper shroud elevations, lifting of a separated shroud due todifferential pressures in the core is resisted by only a small portion of theremaining upper shroud assembly. As such, bypass flow through the gap createdby the separation is sufficient to cause a power/flow mis-match indicationwhich should be observable to the operator during operation. At the shroudlower elevations, the deadweight of the larger portion of the upper shroudassembly can be sufficient to limit lifting of the shroud such that the bypassflow would not be sufficient to be detected.The accident scenarios of primary concern are the main steam line break,recirculation line break and seismic events. The main concern associated withcracks in the upper shroud welds (e.g., H2, H3 in the attached figures) is thesteam line break, since the lifting forces generated may be sufficient toelevate the top guide, possibly affecting lateral support of the fuelassemblies and control rod operation. The main concern associated with cracksin the lower elevations of the core shroud is the postulated recirculationline break. This is because for the lower welds (e.g., H4, H5 in the attachedfigures) the recirculation line break loadings, if large enough, could cause alateral displacement or tipping of the shroud which may affect the ability toinsert the control rods and may result in the opening of a crack that couldallow leakage through the shroud and out through the pipe break. If thisleakage were large enough, it could potentially affect the ability to maintainadequate core cooling, and could affect the ability to shut down the reactorwith the standby liquid control system (SLCS).NRC has developed a probabilistic safety perspective regarding shroudseparation at the lower elevation (Ref. 4) for Dresden, Unit 3 and QuadCities, Unit 1. The assessment estimated the potential contribution to coredamage frequency due to the cracked shroud. Assuming that severe shroudcracking did exist, a large rupture of either a steam or recirculation linewould have to occur to generate loads sufficiently large enough to move the GL 94-03July 25, 1994 shroud. Probabilistic risk assessments categorize such ruptures to be of lowprobability and none has ever actually occurred at an operating nuclear plant.Therefore, the unlikely occurrence of a 360° nearly through-wall crack alongwith a large pipe break would be necessary to pose any incremental risk. Inaddition, for welds in the upper portion of the shroud, through-walldegradation should be detected during normal operation (e.g., by power/flowmis-match or noise monitoring). Finally, the shroud may not move in the mostadverse manner during these events, and there is some likelihood that corecooling and reactor shutdown would be achieved with no adverse consequences.Considering the above evaluations, NRC has made conservative estimates of therisk contribution from shroud cracking and concluded that it does not pose ahigh degree of risk at this time. Although immediate plant shutdown forinspections is not warranted, degradation of the core shroud is an importantsafety consideration warranting further evaluation. The core shroud providesthe important functions of properly directing coolant flow through the core,maintaining the core geometry, and providing a refloodable volume underpostulated accident conditions. The NRC staff therefore considers that 360°cracking of the shroud is a safety concern for the long term based on: (1)potentially exceeding the ASME Code structural margins if the cracks aresufficiently deep and continue to propagate during subsequent operatingcycles; and (2) elimination of a layer of defense-in-depth for plant safety.Therefore, in order to verify compliance with the structural integrityrequirements of 10 CFR 50.55a and to assure that the risk associated with coreshroud cracking remains low, NRC has concluded that it is appropriate for BWRlicensees to implement timely inspections and/or repairs, as appropriate, attheir BWR facilities.Notwithstanding the capability to evaluate the acceptability of cracked coreshrouds for continued operation, the NRC believes that for many of theoperating BWRs that have core shroud materials susceptible to stress corrosioncracking, repairs or additional modifications to inhibit cracking will benecessary to assure structural integrity of the shrouds in the long term.Reguested Licensee ActionsAll addressees are requested to:1. Inspect the core shrouds in their BWR plants no later than the nextscheduled refueling outage;2. Perform a safety analysis supporting continued operation of the facilityuntil inspections are conducted. The safety analysis should consider,but not be limited to the following factors:a. Details of the conditions that would influence the probability ofthe occurrence of cracking and rate of crack growth (e.g.,material types and forms, water chemistry, fluence, carboncontents, welding materials and procedures). | The purpose of this Generic Letter is to request that each addressee: (1)inspect the core shrouds in their BWR plants no later than the next scheduledrefueling outage, and perform an appropriate evaluation and/or repair based onthe results of the inspection; and (2) perform a safety analysis supportingcontinued operation of the facility until inspections are conducted.BackgroundIntergranular stress corrosion cracking (IGSCC) of BWR internal components hasbeen identified as a technical issue of concern by both the NRC staff and theindustry. The core shroud is among the list of internals susceptible toIGSCC. Identification of cracking at the circumferential beltline regionwelds in several plants during 1993 led to the publication of NRC InformationNotice (IN) 93-79, issued on September 30, 1993. Several licensees haverecently inspected their core shrouds during Spring 1994 planned outages andhave identified extensive cracking at the circumferential welds. Theseinspection findings are causing the NRC staff and industry to re-evaluate thesignificance of this issue. Due to the 3600 degree extent of the cracking,and the location at a lower elevation where extensive cracking had not beenpreviously observed (e.g., H5 in the attached figures), the inspections andanalyses performed for Dresden Unit 3 and Quad Cities Unit 1 (Ref. 1, 2) areespecially noteworthy. NRC has issued IN 94-42 on June 7, 1994, andSupplement 1 to IN 94-42 on July 19, 1994, on cracking in the lower region ofthe core shroud found at Dresden Unit 3 and Quad Cities Unit 1. In additionto the core shroud, NRC has an overall concern with cracking of BWR internalsand encourages licensees to work closely with the BWR Owners Group (BWROG) oncoordination of inspections, evaluations and repair options for internalscracking.POP, ocK 050000o 3 P (AOat9407210200 ot 0 d Lo1 4 -i/IC e / ) c Le7J | ||
GL 94-03July 25, 1994 DiscussionEVALUATION OF RECENT INSPECTION EXPERIENCEBy letter dated April 5, 1994 (Ref. 3), the BWROG submitted to NRC genericguidance on the evaluation of BWR core shrouds. This guidance included aninspection strategy that was based on examination of the results of plantinspections up to that time. This inspection strategy was founded on IGSCCsusceptibility rankings and involved focusing the examinations on the uppershroud welds (e.g., H2, H3). Enhanced visual (VT-1) or ultrasonic (UT)methods of portions of the upper shroud welds were recommended for the initialexaminations. However, in light of the recent inspections at Dresden 3 andQuad Cities 1, the BWROG is re-evaluating the applicability of the inspectionguidance. Cracking extending 3600 around the shroud circumference wasobserved at the H5 weld location at both Dresden 3 and Quad Cities 1.Extensive cracking at this location had not been observed previously and wouldnot have been expected based on the BWROG guidance. As all BWR plants havenot performed inspections of their core shrouds, and since the core shroudcracking phenomenon is dependent on operating time and plant specificconditions, additional inspections are necessary to verify that conditionspotentially worse than those already identified do not exist at other plantsand that appropriate corrective actions are taken.SIGNIFICANT PARAMETERS AFFECTING CORE SHROUD CRACKINGThe BWROG has also previously discussed the significant parameters known toaffect the susceptibility of core shrouds to IGSCC (Ref. 3). These parametersinclude, but are not necessarily limited to materials, fabrication andresidual stresses, water chemistry, and fluence. Within and among these broadcategorizations, there exists sufficient variability to make an accurateprediction of IGSCC susceptibility difficult on a generic basis. While theNRC recognizes the usefulness of these categorizations, susceptibility tocracking, or lack thereof, needs to be demonstrated on a plant-specific basis.SIGNIFICANCE OF PART THROUGH-WALL 3600 CRACKINGNRC has assessed the safety significance of part through-wall 3600 core shroudcracking and has concluded that, for the most significant cracks found to date(up to 3600 circumferential extent), the structural margins required by theASME Boiler and Pressure Vessel Code pursuant to Section 50.55a of Title 10 ofthe Code of Federal Regulations [10 CFR 50.55a(g)] were maintained, therebyproviding assurance that the shrouds would have remained intact even underpostulated accident conditions. The ASME Code, Section XI, SubarticleIWB-2500, categories B-N-1 and B-N-2 specify examination and acceptancerequirements for reactor internals and core support structures, including thecore shroud. Paragraph IWB-3520 is referenced as the acceptance standard forintegrally welded core support structures and reactor interior attachments.By letter dated July 13, 1994 (Ref. 4), the BWROG submitted a response toprevious NRC staff questions regarding the susceptibility of BWRs to safety-significant shroud cracks. In this response, the BWROG provided an evaluation GL 94-03July 25, 1994 of the cracking that has been observed in plants which have inspected theirshrouds. The plants which have experienced the most extensive cracking haveoperated for longer than 8 years and had moderate to high coolant conductivityover the first 5 cycles of operation. The BWROG evaluation indicates that thestructural margins for plants most susceptible to cracking would be maintainedfor at least one more cycle of additional operation at current conductivitylevels. However, the BWROG notes that the uncertainties in the assumptionslead to the conclusion that while development of cracks that would not satisfythe ASME Code factors of safety is unlikely, such an occurrence cannot beruled out. Part of the purpose of this generic letter is to ascertain thelikelihood of such an occurrence for each BWR plant and to take appropriatecorrective action(s).SAFETY SIGNIFICANCE OF POSTULATED ACCIDENTS WITH 3600 THROUGH-WALL CRACKSIn order to assess the significance of potential cracking worse than thatobserved to date, NRC has evaluated the safety implications of a postulated3600 circumferential separation of the shroud for which the ASME Code safetymargins are clearly not met. Based on this evaluation, NRC has determinedthat 360° through-wall cracking of the core shroud may not be identified undernormal operating conditions, depending on the elevation of the cracking in theshroud. At the upper shroud elevations, lifting of a separated shroud due todifferential pressures in the core is resisted by only a small portion of theremaining upper shroud assembly. As such, bypass flow through the gap createdby the separation is sufficient to cause a power/flow mis-match indicationwhich should be observable to the operator during operation. At the shroudlower elevations, the deadweight of the larger portion of the upper shroudassembly can be sufficient to limit lifting of the shroud such that the bypassflow would not be sufficient to be detected.The accident scenarios of primary concern are the main steam line break,recirculation line break and seismic events. The main concern associated withcracks in the upper shroud welds (e.g., H2, H3 in the attached figures) is thesteam line break, since the lifting forces generated may be sufficient toelevate the top guide, possibly affecting lateral support of the fuelassemblies and control rod operation. The main concern associated with cracksin the lower elevations of the core shroud is the postulated recirculationline break. This is because for the lower welds (e.g., H4, H5 in the attachedfigures) the recirculation line break loadings, if large enough, could cause alateral displacement or tipping of the shroud which may affect the ability toinsert the control rods and may result in the opening of a crack that couldallow leakage through the shroud and out through the pipe break. If thisleakage were large enough, it could potentially affect the ability to maintainadequate core cooling, and could affect the ability to shut down the reactorwith the standby liquid control system (SLCS).NRC has developed a probabilistic safety perspective regarding shroudseparation at the lower elevation (Ref. 4) for Dresden, Unit 3 and QuadCities, Unit 1. The assessment estimated the potential contribution to coredamage frequency due to the cracked shroud. Assuming that severe shroudcracking did exist, a large rupture of either a steam or recirculation linewould have to occur to generate loads sufficiently large enough to move the GL 94-03July 25, 1994 shroud. Probabilistic risk assessments categorize such ruptures to be of lowprobability and none has ever actually occurred at an operating nuclear plant.Therefore, the unlikely occurrence of a 360° nearly through-wall crack alongwith a large pipe break would be necessary to pose any incremental risk. Inaddition, for welds in the upper portion of the shroud, through-walldegradation should be detected during normal operation (e.g., by power/flowmis-match or noise monitoring). Finally, the shroud may not move in the mostadverse manner during these events, and there is some likelihood that corecooling and reactor shutdown would be achieved with no adverse consequences.Considering the above evaluations, NRC has made conservative estimates of therisk contribution from shroud cracking and concluded that it does not pose ahigh degree of risk at this time. Although immediate plant shutdown forinspections is not warranted, degradation of the core shroud is an importantsafety consideration warranting further evaluation. The core shroud providesthe important functions of properly directing coolant flow through the core,maintaining the core geometry, and providing a refloodable volume underpostulated accident conditions. The NRC staff therefore considers that 360°cracking of the shroud is a safety concern for the long term based on: (1)potentially exceeding the ASME Code structural margins if the cracks aresufficiently deep and continue to propagate during subsequent operatingcycles; and (2) elimination of a layer of defense-in-depth for plant safety.Therefore, in order to verify compliance with the structural integrityrequirements of 10 CFR 50.55a and to assure that the risk associated with coreshroud cracking remains low, NRC has concluded that it is appropriate for BWRlicensees to implement timely inspections and/or repairs, as appropriate, attheir BWR facilities.Notwithstanding the capability to evaluate the acceptability of cracked coreshrouds for continued operation, the NRC believes that for many of theoperating BWRs that have core shroud materials susceptible to stress corrosioncracking, repairs or additional modifications to inhibit cracking will benecessary to assure structural integrity of the shrouds in the long term.Reguested Licensee ActionsAll addressees are requested to:1. Inspect the core shrouds in their BWR plants no later than the nextscheduled refueling outage;2. Perform a safety analysis supporting continued operation of the facilityuntil inspections are conducted. The safety analysis should consider,but not be limited to the following factors:a. Details of the conditions that would influence the probability ofthe occurrence of cracking and rate of crack growth (e.g.,material types and forms, water chemistry, fluence, carboncontents, welding materials and procedures). | |||
GL 94-03July 25, 1994 b. A plant-specific assessment accounting for uncertainties in theamount of cracking, which should include but not be limited to,the following:(1) An assessment of the shroud response to the structuralloadings resulting from design basis events (e.g., steamline break, recirculation line break). If asymmetric loadscan affect the shroud response, these should also beconsidered.(2) An assessment of the ability of plant safety features toperform their function considering the shroud response tostructural loadings (e.g., control rod insertion, ECCSinjection).3. Develop an inspection plan which addresses: (a) all shroud welds (fromsupport attachments to the vessel to the top of the shroud) and/orprovides a justification for elimination of particular welds fromconsideration; and (b) examination methods with appropriateconsideration given to use of the best available technology and industryinspection experience (e.g., enhanced VT-1 visual inspections, optimizedUT techniques). Standard methods for inspection of core supportstructures as specified by the ASME Code, Section XI, have been shown tobe inadequate for consistent detection of IGSCC in core shrouds.4. Develop plans for evaluation and/or repair of the core shroud.5. Work closely with the BWROG on coordination of inspections, evaluationsand repair options for all BWR internals susceptible to IGSCC.Reporting RequirementsPursuant to Section 182a of the Atomic Energy Act of 1954, as amended, and10 CFR 50.54(f), each holder of an operating license for a BWR except Big RockPoint shall submit, under oath or affirmation, the following written responseto this generic letter:1. Within 30 days from the date of this generic letter:(a) A schedule for inspection of the core shroud.(b) A safety analysis, including a plant-specific safety assessment,as appropriate, supporting continued operation of the facilityuntil inspections are conducted.(c) A drawing or drawings of the core shroud configuration showingdetails of the core shroud geometry (e.g., support configurationsfor the lower core support plate and the top guide, weld locationsand configurations).(d) A history of shroud inspections for the plant should be providedaddressing date, scope, methods and results, if applicable. | |||
GL 94-03July 25, 1994 . No later than 3 months prior to performing the core shroud inspections(If the inspections are scheduled to begin in less than 3 months fromthe receipt of this letter, the licensee should contact their NRCproject manager to establish a schedule for providing the followinginformation):(a) The inspection plan requested above in item 3 of RequestedActions.(b) Plans for evaluation and/or repair of the core shroud based on theinspection results.3. Within 30 days from the completion of the inspection, provide theresults of the inspection.The addressee should indicate whether or not the actions requested above willbe implemented in the 30 day response. If an addressee chooses not to takethe requested actions, a description should be provided of any proposedalternative course of action(s), the schedule for completing the alternativecourse of action (if applicable), and the safety basis for determining theacceptability of the planned alternative course of action(s).NRC recognizes that some plant(s) may have already conducted inspectionsand/or performed repairs. However, as the inspection scope and details of themethods employed should reflect cumulative experience to date, as appropriate,this request applies to all BWRs with the exception of Big Rock Point.NRC is also aware that the BWROG is currently developing documents withrevised inspection and flaw assessment guidelines and specifications forrepair options. The response should indicate whether it is intended to followthe guidance developed for this issue by the BWROG. Reference to these andother relevant generic documents developed by the BWROG are acceptable, andencouraged, as part of the response, as long as the referenced documents havebeen officially submitted to NRC. However, as described previously,additional plant-specific information is required to establish thejustification for continued operation.Address these required written reports to the U.S. Nuclear RegulatoryCommission, ATTN: Document Control Desk, Washington, D.C. 20555, under oathor affirmation under the provisions of Section 182a, Atomic Energy Act of1954, as amended, and 10 CFR 50.54(f). In addition, submit a copy to theappropriate regional administrator. | |||
GL 94-03July 25, 1994 Related Generic CommunicationsNRC Information Notice 94-42, Supplement 1, "Cracking In The Lower Region ofthe Core Shroud In Boiling Water Reactors," issued on July 19, 1994.NRC Information Notice 94-42, "Cracking In The Lower Region of the Core ShroudIn Boiling Water Reactors," issued on June 7, 1994.NRC Information Notice 93-79, "Core Shroud Cracking at Beltline Region Weldsin Boiling Water Reactors," issued on September 30, 1993. | |||
==Backfit Discussion== | ==Backfit Discussion== | ||
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==Paperwork Reduction Act Statement== | ==Paperwork Reduction Act Statement== | ||
The information collections contained in this request are covered by theOffice of Management and Budget clearance number 3150-0011, which expiresJuly 31, 1997. The public reporting burden for this collection of informationis estimated to average 350 hours per response, including the time forreviewing instructions, searching existing data sources, gathering andmaintaining the data needs, and completing and reviewing the collection ofinformation. Send comments regarding this burden estimate or any other aspectof this collection of information, including suggestions for reducing thisburden, to the Information and Records Management Branch, (T-6 F33),U.S. Nuclear Regulatory Commission, Washington, D.C., 20555, and to the DeskOfficer, Office of Information and Regulatory Affairs, NEOB-10202,(3150-0011), Office of Management and Budget, Washington, D.C. 20503.Compliance with the following request for information is voluntary. Theinformation would assist the NRC in evaluating the cost of complying with thisgeneric letter.(1) the licensee staff time and costs to perform requested record reviewsand developing plans for inspections; GL 94-03July 25, 1994 (2) the licensee staff time and costs to prepare the requested reports anddocumentation;(3) the additional short-term costs incurred as a result of the inspectionfindings such as the cost of the corrective actions or the costs of downtime; and(4) an estimate of the additional long-term costs that will be incurred as aresult of implementing commitments such as the estimated costs ofconducting future inspections and repairs.If you have any questions about this matter, please contact the technicalcontact listed below or the appropriate NRR project manager.>DAssociat rector for ProjectsOffice o Nuclear Reactor RegulationTechnical contact: Edwin M. Hackett, NRR(301) 504-2751Amy E. Cubbage, NRR(301) 504-2875Lead Project Manager: Donald S. Brinkman, NRR(301) 504- | The information collections contained in this request are covered by theOffice of Management and Budget clearance number 3150-0011, which expiresJuly 31, 1997. The public reporting burden for this collection of informationis estimated to average 350 hours per response, including the time forreviewing instructions, searching existing data sources, gathering andmaintaining the data needs, and completing and reviewing the collection ofinformation. Send comments regarding this burden estimate or any other aspectof this collection of information, including suggestions for reducing thisburden, to the Information and Records Management Branch, (T-6 F33),U.S. Nuclear Regulatory Commission, Washington, D.C., 20555, and to the DeskOfficer, Office of Information and Regulatory Affairs, NEOB-10202,(3150-0011), Office of Management and Budget, Washington, D.C. 20503.Compliance with the following request for information is voluntary. Theinformation would assist the NRC in evaluating the cost of complying with thisgeneric letter.(1) the licensee staff time and costs to perform requested record reviewsand developing plans for inspections; | ||
GL 94-03July 25, 1994 (2) the licensee staff time and costs to prepare the requested reports anddocumentation;(3) the additional short-term costs incurred as a result of the inspectionfindings such as the cost of the corrective actions or the costs of downtime; and(4) an estimate of the additional long-term costs that will be incurred as aresult of implementing commitments such as the estimated costs ofconducting future inspections and repairs.If you have any questions about this matter, please contact the technicalcontact listed below or the appropriate NRR project manager.>DAssociat rector for ProjectsOffice o Nuclear Reactor RegulationTechnical contact: Edwin M. Hackett, NRR(301) 504-2751Amy E. Cubbage, NRR(301) 504-2875Lead Project Manager: Donald S. Brinkman, NRR(301) 504-1409Attachments:1. Figure 1 -Core Shroud Structural Confirguration2. Figure 2 -Core Shroud Weld Locations3. References4. List of Recently Issued NRC Generic Letters | |||
"_iL 94-03July 25, 1994 (2) the licensee staff time and costs to prepare the requested reports anddocumentation;(3) the additional short-term costs incurred as a result of the inspectionfindings such as the cost of the corrective actions or the costs of downtime; and(4) an estimate of the additional long-term costs that will be incurred as aresult of implementing commitments such as the estimated costs ofconducting future inspections and repairs.If you have any questions about this matter, please contact the technicalcontact listed below or the appropriate NRR project manager.oyi, ila signed byRoy P. almmermanAssociate Director for ProjectsOffice of Nuclear Reactor RegulationTechnical contact:Edwin M. Hackett, NRR(301) 504-2751Amy E. Cubbage, NRR(301) 504-2875Lead Project Manager:Donald S. Brinkman, NRR(301) 504-1409Attachments:1. Figure 1 -Core Shroud Structural Confirguration2. Figure 2 -Beta4-1 of-Weld-Locations HSand 864in-thce Dre4dea. Unit '1 "ore Shreu43. References4. List of Recently Issued NRC Generic LetterstdLIA-AOC OGCB,^D99S EMCB:DE jDEB:DDSSA(ONRR,l TEOFC EMCB D JEMCBE :DERRVNAME EHackett n MViDATE 7/25/94 7/25/94 7/25/94 7/25/94 7/25/94 7/22194', -..nrA M 1NPn~rA nT-MPROGNrMBORSDORS:NRR.ADPR:NRR%Pn I %,W-% IT , IV,. -- "rA -.-._~ '__._. .^ .._.NAME JGX Ad 'aklii X r letGreRZimmem5DATE 7/22/94 _ 7/25/94 7/25/94 _ 7/25/94 7/25/94DOCUMENT NAME: 94-03.GL | |||
1. Figure 1 -Core Shroud Structural Confirguration2. Figure 2 -Core Shroud Weld Locations3. References4. List of Recently Issued NRC Generic Letters | I ~Attachment 1GL 94-03July 25, 1994 FIGURE 1CORE SHORUD STRUCTURAL CONFIGURATIONSHROUD HEAD AND SEPARATORS | ||
"_iL 94-03July 25, 1994 (2) the licensee staff time and costs to prepare the requested reports anddocumentation;(3) the additional short-term costs incurred as a result of the inspectionfindings such as the cost of the corrective actions or the costs of downtime; and(4) an estimate of the additional long-term costs that will be incurred as aresult of implementing commitments such as the estimated costs ofconducting future inspections and repairs.If you have any questions about this matter, please contact the technicalcontact listed below or the appropriate NRR project manager.oyi, ila signed byRoy P. almmermanAssociate Director for ProjectsOffice of Nuclear Reactor RegulationTechnical contact:Edwin M. Hackett, NRR(301) 504-2751Amy E. Cubbage, NRR(301) 504-2875Lead Project Manager:Donald S. Brinkman, NRR(301) 504- | ll Attachbent 2GL 94-03July 25, 1994 FIGURE 2-CORE SHROUD WELD LOCATIONS | ||
Attachment 3GL 94-03July 25, 1994 References[1] Letter from M.D. Lyster (Commonwealth Edison) to W.T. Russell(NRC), "Analytical Evaluation of Core Shroud Cracking Identifiedat Dresden Nuclear Power Station Unit 3, NRC Docket No. 50-249,"June 13, 1994.[2] Letter from M.D. Lyster (Commonwealth Edison) to W.T. Russell(NRC), "Analytical Evaluation of Core Shroud Cracking Identifiedat Quad Cities Nuclear Power Station Unit 1, NRC Docket No.50-254," June 13, 1994.(3] Letter from L.A. England, BWROG to USNRC, "Transmittal of BWR CoreShroud Evaluation," GE-NE-523-148-1193, April 5, 1994.[4] Letter from R.A. Pinelli to USNRC, "Response to NRC Request forShroud Information," GE-NE-523-A107P-0794, July 13, 1994. GEPROPRIETARY.[5] NRC Safety Evaluation by the Office of Nuclear Reactor RegulationRelated to Core Shroud Cracking, Commonwealth Edison Company andIowa-Illinois Gas and Electric Company, Dresden Nuclear PowerStation, Unit 3 and Quad Cities Nuclear Power Station, Unit 1,Docket Nos. 50-249 and 50-254, July, 1994. | |||
K-'Attachment 4GL 94-03July 25, 1994 LIST OF RECENTLY ISSUED GENERIC LETTERSGenericI £^4++nDate ofT cc ismnrC.s ak4s-Tectnord TnL.I6IE:VI IUU.l3~.L ,a-u1- -a --94-0294-0186-10,SUPP. 189-10,SUPP. 693-0893-07LONG-TERM SOLUTIONS ANDUPGRADE OF INTERIMOPERATING RECOMMENDATIONSFOR THERMAL-HYDRAULICINSTABILITIES IN BOILINGWATER REACTORSREMOVAL OF ACCELERATEDTESTING AND SPECIAL RE-PORTING REQUIREMENTS FOREMERGENCY DIESEL GENERATORSFIRE ENDURANCE TESTACCEPTANCE CRITERIA FORFIRE BARRIER SYSTEMS USEDTO SEPARATE REDUNDANTSAFE SHUTDOWN TRAINS WITHINTHE SAME FIRE AREA (SUPP. 1 TOGL 86-10, "IMPLEMENTATION OFFIRE PROTECTION REQUIREMENTS")INFORMATION ON SCHEDULEAND GROUPING, AND STAFFRESPONSES TO ADDITIONALPUBLIC QUESTIONSRELOCATION OF TECHNICALSPECIFICATION TABLES OFOF INSTRUMENT RESPONSETIME LIMITSMODIFICATION OF THE TECH-NICAL SPECIFICATION ADMINI-STRATIVE CONTROL REQUIRE-MENTS FOR EMERGENCY ANDSECURITY PLANS07/11/9405/31/9403/25/9403/08/9412/29/9312/28/93ALL HOLDERS OF OLs FORBOILING WATER REACTORSEXCEPT BIG ROCK POINTALL HOLDERS OF OLs FORNPRsALL HOLDERS OF OLs OR CPsFOR NPRsALL LICENSEES OFOPERATING NUCLEARPOWER PLANTS AND HOLDERSOF CONSTRUCTION PERMITSFOR NPRsALL HOLDERS OF OLs FORNPRs-ALL--HOLDERS VF OLs ORCPs FOR NPRsOL = OPERATING LICENSECP = CONSTRUCTION PERMITNPR = NUCLEAR POWER REACTORS | |||
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- -II!UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF NUCLEAR REACTOR REGULATIONWASHINGTON, D.C. 20555July 25, 1994NRC GENERIC LETTER 94-03: INTERGRANULAR STRESS CORROSION CRACKING OF CORESHROUDS IN BOILING WATER REACTORS
Addressees
All holders of operating licenses or construction permits for boiling waterreactors (BWRs) except for Big Rock Point, which does not have a core shroud.
Purpose
The purpose of this Generic Letter is to request that each addressee: (1)inspect the core shrouds in their BWR plants no later than the next scheduledrefueling outage, and perform an appropriate evaluation and/or repair based onthe results of the inspection; and (2) perform a safety analysis supportingcontinued operation of the facility until inspections are conducted.BackgroundIntergranular stress corrosion cracking (IGSCC) of BWR internal components hasbeen identified as a technical issue of concern by both the NRC staff and theindustry. The core shroud is among the list of internals susceptible toIGSCC. Identification of cracking at the circumferential beltline regionwelds in several plants during 1993 led to the publication of NRC InformationNotice (IN) 93-79, issued on September 30, 1993. Several licensees haverecently inspected their core shrouds during Spring 1994 planned outages andhave identified extensive cracking at the circumferential welds. Theseinspection findings are causing the NRC staff and industry to re-evaluate thesignificance of this issue. Due to the 3600 degree extent of the cracking,and the location at a lower elevation where extensive cracking had not beenpreviously observed (e.g., H5 in the attached figures), the inspections andanalyses performed for Dresden Unit 3 and Quad Cities Unit 1 (Ref. 1, 2) areespecially noteworthy. NRC has issued IN 94-42 on June 7, 1994, andSupplement 1 to IN 94-42 on July 19, 1994, on cracking in the lower region ofthe core shroud found at Dresden Unit 3 and Quad Cities Unit 1. In additionto the core shroud, NRC has an overall concern with cracking of BWR internalsand encourages licensees to work closely with the BWR Owners Group (BWROG) oncoordination of inspections, evaluations and repair options for internalscracking.POP, ocK 050000o 3 P (AOat9407210200 ot 0 d Lo1 4 -i/IC e / ) c Le7J
GL 94-03July 25, 1994 DiscussionEVALUATION OF RECENT INSPECTION EXPERIENCEBy letter dated April 5, 1994 (Ref. 3), the BWROG submitted to NRC genericguidance on the evaluation of BWR core shrouds. This guidance included aninspection strategy that was based on examination of the results of plantinspections up to that time. This inspection strategy was founded on IGSCCsusceptibility rankings and involved focusing the examinations on the uppershroud welds (e.g., H2, H3). Enhanced visual (VT-1) or ultrasonic (UT)methods of portions of the upper shroud welds were recommended for the initialexaminations. However, in light of the recent inspections at Dresden 3 andQuad Cities 1, the BWROG is re-evaluating the applicability of the inspectionguidance. Cracking extending 3600 around the shroud circumference wasobserved at the H5 weld location at both Dresden 3 and Quad Cities 1.Extensive cracking at this location had not been observed previously and wouldnot have been expected based on the BWROG guidance. As all BWR plants havenot performed inspections of their core shrouds, and since the core shroudcracking phenomenon is dependent on operating time and plant specificconditions, additional inspections are necessary to verify that conditionspotentially worse than those already identified do not exist at other plantsand that appropriate corrective actions are taken.SIGNIFICANT PARAMETERS AFFECTING CORE SHROUD CRACKINGThe BWROG has also previously discussed the significant parameters known toaffect the susceptibility of core shrouds to IGSCC (Ref. 3). These parametersinclude, but are not necessarily limited to materials, fabrication andresidual stresses, water chemistry, and fluence. Within and among these broadcategorizations, there exists sufficient variability to make an accurateprediction of IGSCC susceptibility difficult on a generic basis. While theNRC recognizes the usefulness of these categorizations, susceptibility tocracking, or lack thereof, needs to be demonstrated on a plant-specific basis.SIGNIFICANCE OF PART THROUGH-WALL 3600 CRACKINGNRC has assessed the safety significance of part through-wall 3600 core shroudcracking and has concluded that, for the most significant cracks found to date(up to 3600 circumferential extent), the structural margins required by theASME Boiler and Pressure Vessel Code pursuant to Section 50.55a of Title 10 ofthe Code of Federal Regulations [10 CFR 50.55a(g)] were maintained, therebyproviding assurance that the shrouds would have remained intact even underpostulated accident conditions. The ASME Code,Section XI, SubarticleIWB-2500, categories B-N-1 and B-N-2 specify examination and acceptancerequirements for reactor internals and core support structures, including thecore shroud. Paragraph IWB-3520 is referenced as the acceptance standard forintegrally welded core support structures and reactor interior attachments.By letter dated July 13, 1994 (Ref. 4), the BWROG submitted a response toprevious NRC staff questions regarding the susceptibility of BWRs to safety-significant shroud cracks. In this response, the BWROG provided an evaluation GL 94-03July 25, 1994 of the cracking that has been observed in plants which have inspected theirshrouds. The plants which have experienced the most extensive cracking haveoperated for longer than 8 years and had moderate to high coolant conductivityover the first 5 cycles of operation. The BWROG evaluation indicates that thestructural margins for plants most susceptible to cracking would be maintainedfor at least one more cycle of additional operation at current conductivitylevels. However, the BWROG notes that the uncertainties in the assumptionslead to the conclusion that while development of cracks that would not satisfythe ASME Code factors of safety is unlikely, such an occurrence cannot beruled out. Part of the purpose of this generic letter is to ascertain thelikelihood of such an occurrence for each BWR plant and to take appropriatecorrective action(s).SAFETY SIGNIFICANCE OF POSTULATED ACCIDENTS WITH 3600 THROUGH-WALL CRACKSIn order to assess the significance of potential cracking worse than thatobserved to date, NRC has evaluated the safety implications of a postulated3600 circumferential separation of the shroud for which the ASME Code safetymargins are clearly not met. Based on this evaluation, NRC has determinedthat 360° through-wall cracking of the core shroud may not be identified undernormal operating conditions, depending on the elevation of the cracking in theshroud. At the upper shroud elevations, lifting of a separated shroud due todifferential pressures in the core is resisted by only a small portion of theremaining upper shroud assembly. As such, bypass flow through the gap createdby the separation is sufficient to cause a power/flow mis-match indicationwhich should be observable to the operator during operation. At the shroudlower elevations, the deadweight of the larger portion of the upper shroudassembly can be sufficient to limit lifting of the shroud such that the bypassflow would not be sufficient to be detected.The accident scenarios of primary concern are the main steam line break,recirculation line break and seismic events. The main concern associated withcracks in the upper shroud welds (e.g., H2, H3 in the attached figures) is thesteam line break, since the lifting forces generated may be sufficient toelevate the top guide, possibly affecting lateral support of the fuelassemblies and control rod operation. The main concern associated with cracksin the lower elevations of the core shroud is the postulated recirculationline break. This is because for the lower welds (e.g., H4, H5 in the attachedfigures) the recirculation line break loadings, if large enough, could cause alateral displacement or tipping of the shroud which may affect the ability toinsert the control rods and may result in the opening of a crack that couldallow leakage through the shroud and out through the pipe break. If thisleakage were large enough, it could potentially affect the ability to maintainadequate core cooling, and could affect the ability to shut down the reactorwith the standby liquid control system (SLCS).NRC has developed a probabilistic safety perspective regarding shroudseparation at the lower elevation (Ref. 4) for Dresden, Unit 3 and QuadCities, Unit 1. The assessment estimated the potential contribution to coredamage frequency due to the cracked shroud. Assuming that severe shroudcracking did exist, a large rupture of either a steam or recirculation linewould have to occur to generate loads sufficiently large enough to move the GL 94-03July 25, 1994 shroud. Probabilistic risk assessments categorize such ruptures to be of lowprobability and none has ever actually occurred at an operating nuclear plant.Therefore, the unlikely occurrence of a 360° nearly through-wall crack alongwith a large pipe break would be necessary to pose any incremental risk. Inaddition, for welds in the upper portion of the shroud, through-walldegradation should be detected during normal operation (e.g., by power/flowmis-match or noise monitoring). Finally, the shroud may not move in the mostadverse manner during these events, and there is some likelihood that corecooling and reactor shutdown would be achieved with no adverse consequences.Considering the above evaluations, NRC has made conservative estimates of therisk contribution from shroud cracking and concluded that it does not pose ahigh degree of risk at this time. Although immediate plant shutdown forinspections is not warranted, degradation of the core shroud is an importantsafety consideration warranting further evaluation. The core shroud providesthe important functions of properly directing coolant flow through the core,maintaining the core geometry, and providing a refloodable volume underpostulated accident conditions. The NRC staff therefore considers that 360°cracking of the shroud is a safety concern for the long term based on: (1)potentially exceeding the ASME Code structural margins if the cracks aresufficiently deep and continue to propagate during subsequent operatingcycles; and (2) elimination of a layer of defense-in-depth for plant safety.Therefore, in order to verify compliance with the structural integrityrequirements of 10 CFR 50.55a and to assure that the risk associated with coreshroud cracking remains low, NRC has concluded that it is appropriate for BWRlicensees to implement timely inspections and/or repairs, as appropriate, attheir BWR facilities.Notwithstanding the capability to evaluate the acceptability of cracked coreshrouds for continued operation, the NRC believes that for many of theoperating BWRs that have core shroud materials susceptible to stress corrosioncracking, repairs or additional modifications to inhibit cracking will benecessary to assure structural integrity of the shrouds in the long term.Reguested Licensee ActionsAll addressees are requested to:1. Inspect the core shrouds in their BWR plants no later than the nextscheduled refueling outage;2. Perform a safety analysis supporting continued operation of the facilityuntil inspections are conducted. The safety analysis should consider,but not be limited to the following factors:a. Details of the conditions that would influence the probability ofthe occurrence of cracking and rate of crack growth (e.g.,material types and forms, water chemistry, fluence, carboncontents, welding materials and procedures).
GL 94-03July 25, 1994 b. A plant-specific assessment accounting for uncertainties in theamount of cracking, which should include but not be limited to,the following:(1) An assessment of the shroud response to the structuralloadings resulting from design basis events (e.g., steamline break, recirculation line break). If asymmetric loadscan affect the shroud response, these should also beconsidered.(2) An assessment of the ability of plant safety features toperform their function considering the shroud response tostructural loadings (e.g., control rod insertion, ECCSinjection).3. Develop an inspection plan which addresses: (a) all shroud welds (fromsupport attachments to the vessel to the top of the shroud) and/orprovides a justification for elimination of particular welds fromconsideration; and (b) examination methods with appropriateconsideration given to use of the best available technology and industryinspection experience (e.g., enhanced VT-1 visual inspections, optimizedUT techniques). Standard methods for inspection of core supportstructures as specified by the ASME Code,Section XI, have been shown tobe inadequate for consistent detection of IGSCC in core shrouds.4. Develop plans for evaluation and/or repair of the core shroud.5. Work closely with the BWROG on coordination of inspections, evaluationsand repair options for all BWR internals susceptible to IGSCC.Reporting RequirementsPursuant to Section 182a of the Atomic Energy Act of 1954, as amended, and10 CFR 50.54(f), each holder of an operating license for a BWR except Big RockPoint shall submit, under oath or affirmation, the following written responseto this generic letter:1. Within 30 days from the date of this generic letter:(a) A schedule for inspection of the core shroud.(b) A safety analysis, including a plant-specific safety assessment,as appropriate, supporting continued operation of the facilityuntil inspections are conducted.(c) A drawing or drawings of the core shroud configuration showingdetails of the core shroud geometry (e.g., support configurationsfor the lower core support plate and the top guide, weld locationsand configurations).(d) A history of shroud inspections for the plant should be providedaddressing date, scope, methods and results, if applicable.
GL 94-03July 25, 1994 . No later than 3 months prior to performing the core shroud inspections(If the inspections are scheduled to begin in less than 3 months fromthe receipt of this letter, the licensee should contact their NRCproject manager to establish a schedule for providing the followinginformation):(a) The inspection plan requested above in item 3 of RequestedActions.(b) Plans for evaluation and/or repair of the core shroud based on theinspection results.3. Within 30 days from the completion of the inspection, provide theresults of the inspection.The addressee should indicate whether or not the actions requested above willbe implemented in the 30 day response. If an addressee chooses not to takethe requested actions, a description should be provided of any proposedalternative course of action(s), the schedule for completing the alternativecourse of action (if applicable), and the safety basis for determining theacceptability of the planned alternative course of action(s).NRC recognizes that some plant(s) may have already conducted inspectionsand/or performed repairs. However, as the inspection scope and details of themethods employed should reflect cumulative experience to date, as appropriate,this request applies to all BWRs with the exception of Big Rock Point.NRC is also aware that the BWROG is currently developing documents withrevised inspection and flaw assessment guidelines and specifications forrepair options. The response should indicate whether it is intended to followthe guidance developed for this issue by the BWROG. Reference to these andother relevant generic documents developed by the BWROG are acceptable, andencouraged, as part of the response, as long as the referenced documents havebeen officially submitted to NRC. However, as described previously,additional plant-specific information is required to establish thejustification for continued operation.Address these required written reports to the U.S. Nuclear RegulatoryCommission, ATTN: Document Control Desk, Washington, D.C. 20555, under oathor affirmation under the provisions of Section 182a, Atomic Energy Act of1954, as amended, and 10 CFR 50.54(f). In addition, submit a copy to theappropriate regional administrator.
GL 94-03July 25, 1994 Related Generic CommunicationsNRC Information Notice 94-42, Supplement 1, "Cracking In The Lower Region ofthe Core Shroud In Boiling Water Reactors," issued on July 19, 1994.NRC Information Notice 94-42, "Cracking In The Lower Region of the Core ShroudIn Boiling Water Reactors," issued on June 7, 1994.NRC Information Notice 93-79, "Core Shroud Cracking at Beltline Region Weldsin Boiling Water Reactors," issued on September 30, 1993.
Backfit Discussion
The actions requested in this generic letter are considered backfits inaccordance with NRC procedures. These backfits are necessary to verify thatthe addressees are in compliance with existing requirements. Therefore, onthe basis of 10 CFR 50.109(a)(4)(i), a full backfit analysis was notperformed. An evaluation was performed in accordance with NRC procedures,including a statement of the objectives of and reasons for the requestedactions and the basis for invoking the compliance exception. A copy of thisevaluation will be made available in the public document room.A notice of opportunity for public comment was not published in the FederalRegister because of the urgent nature of the actions requested by the genericletter.
Paperwork Reduction Act Statement
The information collections contained in this request are covered by theOffice of Management and Budget clearance number 3150-0011, which expiresJuly 31, 1997. The public reporting burden for this collection of informationis estimated to average 350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br /> per response, including the time forreviewing instructions, searching existing data sources, gathering andmaintaining the data needs, and completing and reviewing the collection ofinformation. Send comments regarding this burden estimate or any other aspectof this collection of information, including suggestions for reducing thisburden, to the Information and Records Management Branch, (T-6 F33),U.S. Nuclear Regulatory Commission, Washington, D.C., 20555, and to the DeskOfficer, Office of Information and Regulatory Affairs, NEOB-10202,(3150-0011), Office of Management and Budget, Washington, D.C. 20503.Compliance with the following request for information is voluntary. Theinformation would assist the NRC in evaluating the cost of complying with thisgeneric letter.(1) the licensee staff time and costs to perform requested record reviewsand developing plans for inspections;
GL 94-03July 25, 1994 (2) the licensee staff time and costs to prepare the requested reports anddocumentation;(3) the additional short-term costs incurred as a result of the inspectionfindings such as the cost of the corrective actions or the costs of downtime; and(4) an estimate of the additional long-term costs that will be incurred as aresult of implementing commitments such as the estimated costs ofconducting future inspections and repairs.If you have any questions about this matter, please contact the technicalcontact listed below or the appropriate NRR project manager.>DAssociat rector for ProjectsOffice o Nuclear Reactor RegulationTechnical contact: Edwin M. Hackett, NRR(301) 504-2751Amy E. Cubbage, NRR(301) 504-2875Lead Project Manager: Donald S. Brinkman, NRR(301) 504-1409Attachments:1. Figure 1 -Core Shroud Structural Confirguration2. Figure 2 -Core Shroud Weld Locations3. References4. List of Recently Issued NRC Generic Letters
"_iL 94-03July 25, 1994 (2) the licensee staff time and costs to prepare the requested reports anddocumentation;(3) the additional short-term costs incurred as a result of the inspectionfindings such as the cost of the corrective actions or the costs of downtime; and(4) an estimate of the additional long-term costs that will be incurred as aresult of implementing commitments such as the estimated costs ofconducting future inspections and repairs.If you have any questions about this matter, please contact the technicalcontact listed below or the appropriate NRR project manager.oyi, ila signed byRoy P. almmermanAssociate Director for ProjectsOffice of Nuclear Reactor RegulationTechnical contact:Edwin M. Hackett, NRR(301) 504-2751Amy E. Cubbage, NRR(301) 504-2875Lead Project Manager:Donald S. Brinkman, NRR(301) 504-1409Attachments:1. Figure 1 -Core Shroud Structural Confirguration2. Figure 2 -Beta4-1 of-Weld-Locations HSand 864in-thce Dre4dea. Unit '1 "ore Shreu43. References4. List of Recently Issued NRC Generic LetterstdLIA-AOC OGCB,^D99S EMCB:DE jDEB:DDSSA(ONRR,l TEOFC EMCB D JEMCBE :DERRVNAME EHackett n MViDATE 7/25/94 7/25/94 7/25/94 7/25/94 7/25/94 7/22194', -..nrA M 1NPn~rA nT-MPROGNrMBORSDORS:NRR.ADPR:NRR%Pn I %,W-% IT , IV,. -- "rA -.-._~ '__._. .^ .._.NAME JGX Ad 'aklii X r letGreRZimmem5DATE 7/22/94 _ 7/25/94 7/25/94 _ 7/25/94 7/25/94DOCUMENT NAME: 94-03.GL
I ~Attachment 1GL 94-03July 25, 1994 FIGURE 1CORE SHORUD STRUCTURAL CONFIGURATIONSHROUD HEAD AND SEPARATORS
ll Attachbent 2GL 94-03July 25, 1994 FIGURE 2-CORE SHROUD WELD LOCATIONS
Attachment 3GL 94-03July 25, 1994 References[1] Letter from M.D. Lyster (Commonwealth Edison) to W.T. Russell(NRC), "Analytical Evaluation of Core Shroud Cracking Identifiedat Dresden Nuclear Power Station Unit 3, NRC Docket No. 50-249,"June 13, 1994.[2] Letter from M.D. Lyster (Commonwealth Edison) to W.T. Russell(NRC), "Analytical Evaluation of Core Shroud Cracking Identifiedat Quad Cities Nuclear Power Station Unit 1, NRC Docket No.50-254," June 13, 1994.(3] Letter from L.A. England, BWROG to USNRC, "Transmittal of BWR CoreShroud Evaluation," GE-NE-523-148-1193, April 5, 1994.[4] Letter from R.A. Pinelli to USNRC, "Response to NRC Request forShroud Information," GE-NE-523-A107P-0794, July 13, 1994. GEPROPRIETARY.[5] NRC Safety Evaluation by the Office of Nuclear Reactor RegulationRelated to Core Shroud Cracking, Commonwealth Edison Company andIowa-Illinois Gas and Electric Company, Dresden Nuclear PowerStation, Unit 3 and Quad Cities Nuclear Power Station, Unit 1,Docket Nos. 50-249 and 50-254, July, 1994.
K-'Attachment 4GL 94-03July 25, 1994 LIST OF RECENTLY ISSUED GENERIC LETTERSGenericI £^4++nDate ofT cc ismnrC.s ak4s-Tectnord TnL.I6IE:VI IUU.l3~.L ,a-u1- -a --94-0294-0186-10,SUPP. 189-10,SUPP. 693-0893-07LONG-TERM SOLUTIONS ANDUPGRADE OF INTERIMOPERATING RECOMMENDATIONSFOR THERMAL-HYDRAULICINSTABILITIES IN BOILINGWATER REACTORSREMOVAL OF ACCELERATEDTESTING AND SPECIAL RE-PORTING REQUIREMENTS FOREMERGENCY DIESEL GENERATORSFIRE ENDURANCE TESTACCEPTANCE CRITERIA FORFIRE BARRIER SYSTEMS USEDTO SEPARATE REDUNDANTSAFE SHUTDOWN TRAINS WITHINTHE SAME FIRE AREA (SUPP. 1 TOGL 86-10, "IMPLEMENTATION OFFIRE PROTECTION REQUIREMENTS")INFORMATION ON SCHEDULEAND GROUPING, AND STAFFRESPONSES TO ADDITIONALPUBLIC QUESTIONSRELOCATION OF TECHNICALSPECIFICATION TABLES OFOF INSTRUMENT RESPONSETIME LIMITSMODIFICATION OF THE TECH-NICAL SPECIFICATION ADMINI-STRATIVE CONTROL REQUIRE-MENTS FOR EMERGENCY ANDSECURITY PLANS07/11/9405/31/9403/25/9403/08/9412/29/9312/28/93ALL HOLDERS OF OLs FORBOILING WATER REACTORSEXCEPT BIG ROCK POINTALL HOLDERS OF OLs FORNPRsALL HOLDERS OF OLs OR CPsFOR NPRsALL LICENSEES OFOPERATING NUCLEARPOWER PLANTS AND HOLDERSOF CONSTRUCTION PERMITSFOR NPRsALL HOLDERS OF OLs FORNPRs-ALL--HOLDERS VF OLs ORCPs FOR NPRsOL = OPERATING LICENSECP = CONSTRUCTION PERMITNPR = NUCLEAR POWER REACTORS