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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20195C4271998-11-0606 November 1998 Safety Evaluation Supporting Proposed Ocnpp Mod to Install Core Support Plate Wedges to Structurally Replace Lateral Resistance Provided by Rim Hold Down Bolts for One Operating Cycle ML20154L3051998-10-14014 October 1998 Safety Evaluation Accepting Licensee Request to Defer Insp of 79 Welds from One Fuel Cycle at 17R Outage ML20217A4631998-03-23023 March 1998 Safety Evaluation Accepting Use of Three Heats/Lots of Hot Rolled XM-19 Matl in Core Shroud Repair Assemblies Re Licenses DPR-16 & DPR-59,respectively ML20199E4561997-11-13013 November 1997 Safety Evaluation Accepting Ampacity Derating Analysis in Response to NRC RAI Re GL-92-08, Thermo-Lag 330-1 Fire Barriers, for Plant ML20149F9961997-07-18018 July 1997 Safety Evaluation Re Gpu Nuclear Operational Quality Assurance Plan,Rev 10 for Three Mile Island Nuclear Generating Station,Unit 1 & Oyster Creek Nuclear Generating Station ML20133E1681997-01-0707 January 1997 Safety Evaluation Re Third 10-yr Interval ISI Plan,Rev 1 to Relief Request R11 for Plant.Proposed Alternative to ASME Requirements Authorized ML20128L1601996-10-0303 October 1996 Safety Evaluation Accepting Third 10-yr Interval Inservice Insp Plan Request for Relief R15 ML20128F4791996-10-0101 October 1996 Safety Evaluation Accepting Rev to Inservice Testing Program Re Leakage Testing of Containment Isolation Valves ML20056H2651993-08-24024 August 1993 SE Re Inservice Testing Program Requests for Relief ML20056E0911993-08-0404 August 1993 SE Re Util 930614 Response to Bulletin 93-03, Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in Bwrs. Util Justification for Not Implementing Addl Short Term Actions Acceptable ML20128P4651993-02-18018 February 1993 Safety Evaluation Accepting Util Justification for Cancelling Commitment on Five Plant Control Room Human Engineering Discrepancies Re Relocation of Shift Supervisor Ofc ML20128F1361993-02-0505 February 1993 Safety Evaluation Re Leak on Core Spray in-vessel Annulus Piping.Plant Can Be Safely Operated for One Fuel Cycle W/O Repairing Observed Leak at Listed 1/4-inch Fillet Weld ML20125C3901992-12-0707 December 1992 Safety Evaluation Re Upper Reactor Bldg & Nonsafety Architectural Components Subjected to tornado-wind Loading ML20127P2251992-11-23023 November 1992 Safety Evaluation Accepting Response to SBO Rule ML20058A7621990-10-18018 October 1990 SE Accepting Util Insp & Repairs for Igscc,Per Generic Ltr 88-01 ML20246M7561989-08-30030 August 1989 Safety Evaluation Accepting Util 890624 Response to Integrated Plant Safety Assessment Section 4.11, Seismic Design Consideration, Items 4.11(1), Piping Sys & 4.11(3), Electrical Equipment ML20247A7061989-07-17017 July 1989 Safety Evaluation Supporting Generic Ltr 83-28,Item 4.5.2 Re Periodic on-line Testing of Reactor Trip Sys ML20245C1711989-06-15015 June 1989 Safety Evaluation Re Generic Ltr 83-28,Item 4.5.3 Re Reactor Trip Sys Reliability for All Domestic Operating Reactors ML20244C8501989-06-12012 June 1989 Safety Evaluation Accepting Util Response to Generic Ltr 83-28,Item 4.5.3 Re on-line Testing for Reactor Trip Sys NUREG-1000, Safety Evaluation Accepting Util 831114 & 851023 Responses to Generic Ltr 83-28,Item 2.1 (Part 1),confirming That All Applicable Components Identified as safety-related1989-04-0303 April 1989 Safety Evaluation Accepting Util 831114 & 851023 Responses to Generic Ltr 83-28,Item 2.1 (Part 1),confirming That All Applicable Components Identified as safety-related ML20235X1281989-03-0606 March 1989 Safety Evaluation Supporting Licensee Request to Extend Second 10-yr Inservice Insp & Testing Interval Until 911014 ML20205P5291988-11-0404 November 1988 Safety Evaluation Supporting Util Submittal Re Compliance W/Atws Rule 10CFR50.62 Concerning Alternate Rod Injection & Recirculation Pump Trip Sys ML20207L0831988-10-11011 October 1988 Safety Evaluation Re Systematic Evaluation of Ipsar Topic VII-1.A, Isolation of Reactor Protection Sys from Non-Safety Sys. Issue Resolved IR 05000219/19840311988-03-0404 March 1988 Safety Evaluation Concluding Util Compliance W/Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1,3.2.2 & 4.5.1,based on Util 831114,850809,0509,1023 & 871223 Descriptions of Planned & Completed Actions & Insp Rept 50-219/84-31 on 841126-30 ML20149L6381988-02-18018 February 1988 Safety Evaluation Supporting Proposed Standby Liquid Control Sys Operation ML20236Q5311987-11-0606 November 1987 Safety Evaluation Re Safety Limit Violation on 870911.Safety Significance Considered Low Due to Adequate Core Cooling & Low Decay Heat Levels.Basis for Conclusion Elaborated in Encl Insp Rept 50-219/87-29 ML20235E0661987-09-22022 September 1987 Safety Evaluation Re Corrosion of Drywell Shell.Licensee Meets NRC Requirements as Specified in 861224 Safety Evaluation ML20207S2311987-03-13013 March 1987 Safety Evaluation Re Generic Ltr 84-09 Re Hydrogen Recombiner Capability.Licensee Should Provide Nitrogen Containment Atmosphere Dilution Sys Capable of Isolating Air from Containment Whenever Isolation Signal Occurs ML20212M3561987-03-0303 March 1987 Safety Evaluation Rejecting Util Schedule for Completion of Mark I Containment Program Re Vacuum Breaker Integrity. Required Vacuum Breaker Parts Should Be Replaced Prior to Startup from Cycle 12 Refueling Outage ML20207M2081986-12-29029 December 1986 Safety Evaluation Re Corrosion of Drywell Shell Line Break in Isolation Condenser.Operation of Plant for Cycle 12 Safe W/Drywell Steel Plates Not Fully in Conformance W/Fsar.Util Shall Provide Plans for Mitigating Problem by 870630 ML20215F3681986-12-15015 December 1986 Safety Evaluation Supporting Util Response to IE Bulletin 86-002 Re Static O-Ring Differential Pressure Switches ML20211M3911986-11-28028 November 1986 Safety Evaluation Re Deferment of NRC Required Mods from Cycle 11R Outage,Including SPDS Implementation,Isolation Condenser Makeup Pump,Intake Canal Level Instrumentation & Masonry Wall Mods.Request Acceptable ML20214A5571986-11-14014 November 1986 Safety Evaluation Re Insp & Repair of Reactor Coolant Piping Sys & Supporting Return to Operation for Next Operating Cycle ML20213G5551986-10-31031 October 1986 Safety Evaluation Supporting Util 860903 Submittal Re Insp of Core Spray Sparger Sys in Current Cycle 11 Refueling Outage.Licensee Satisfied Requirements of License Condition 2.C.(7) for Current Cycle 11 Refueling Outage NUREG-0822, Safety Evaluation Finding Integrated Plant SAR (NUREG-0822), Section 4.12 Complete W/Exception of Three Issues,Including Evaluation of Drywell for Concrete Subj to High Temps & Thermal Transients1986-10-29029 October 1986 Safety Evaluation Finding Integrated Plant SAR (NUREG-0822), Section 4.12 Complete W/Exception of Three Issues,Including Evaluation of Drywell for Concrete Subj to High Temps & Thermal Transients ML20197B8801986-10-10010 October 1986 Safety Evaluation Supporting Util 850924 Proposal Not to Replace Existing Containment Purge & Vent Isolation Valves W/New Clow Valves ML20215E7261986-10-0101 October 1986 Safety Evaluation Supporting Cancellation of Torus Pool Temp Indication & Thermal Mixing Mods ML20204F9721986-07-28028 July 1986 Corrected Page 3 to 860522 Safety Evaluation Re Integrated Plant Safety Assessment Rept,Section 4.38, Locas. Correction Concerns Licensee Maint to Repack Four MSIVs Each Refueling Outage ML20210K2481986-04-22022 April 1986 Safety Evaluation Supporting Util 831114 Response to Generic Ltr 83-28,Item 1.2, Post-Trip Review,Data & Info Capability ML20141D4261986-03-31031 March 1986 Safety Evaluation Re Generic Ltr 83-28,Items 3.1.3 & 3.2.3, Post-Maint Testing (Program Description & Procedure). Tech Specs Incorporating post-maint Testing of Reactor Protection Sys Components Should Be Proposed ML20141N0451986-02-24024 February 1986 Safety Evaluation Supporting Deferment of Feedwater Nozzle Insps Until Cycle 12R Outage Scheduled for 1988 ML20205H7981986-01-21021 January 1986 Safety Evaluation Re Venting & Purging Containment During Plant Operation.Rationale Utilized for Design & Operation of Purge/Vent Sys Acceptable ML20136A9351985-12-23023 December 1985 Safety Evaluation Accepting Licensee Response to Items 2(b) & 3 of IE Bulletin 80-11, Masonry Wall Design. Licensee 850726 Request Re Wall Mods to Be Completed During Operating Cycle 11 Not Addressed ML20137Q7621985-11-22022 November 1985 Safety Evaluation Supporting Determination That Max 150 F Drywell Temp Acceptable for Containment Safety,Drywell Steel & Concrete Structural Components & Normal Plant Operations. Tech Specs Re Limiting Factors Requested ML20128H5241985-05-28028 May 1985 Safety Evaluation Supporting Environ Qualification of Electric Equipment Important to Safety.Proposed Resolutions for Environ Qualification Deficiencies Acceptable ML20126K1221981-03-29029 March 1981 Safety Evaluation Supporting Amend 54 to License DPR-16 1998-03-23
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217K4451999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Oyster Creek Nuclear Generating Station.With ML20211P6731999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Oyster Creek Nuclear Generating Station.With ML20211A7051999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Oyster Creek Nuclear Station.With ML20209G0631999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Oyster Creek Nuclear Generating Station.With ML20212H5491999-06-18018 June 1999 Non-proprietary Rev 4 to HI-981983, Licensing Rept for Storage Capacity Expansion of Oyster Creek Spent Fuel Pool ML20195E7961999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Oyster Creek Nuclear Generating Station.With ML20206N7431999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Oyster Creek Nuclear Generating Station.With ML20205P5401999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Oyster Creek Nuclear Generating Station.With ML20204C8201999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Oyster Creek Nuclear Generating Station.With ML20199E4671998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Oyster Creek Nuclear Generating Station.With ML20195E8321998-12-31031 December 1998 10CFR50.59(b) Rept of Changes to Oyster Creek Sys & Procedures, for Period of June 1997 to Dec 1998.With ML20198D2091998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Oyster Creek Nuclear Generating Station.With ML20195J8591998-11-12012 November 1998 Rev 11 to 1000-PLN-7200.01, Gpu Nuclear Operational QA Plan ML20195C4271998-11-0606 November 1998 Safety Evaluation Supporting Proposed Ocnpp Mod to Install Core Support Plate Wedges to Structurally Replace Lateral Resistance Provided by Rim Hold Down Bolts for One Operating Cycle ML20155J3021998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Oyster Creek Nuclear Generating Station.With ML20154R4981998-10-20020 October 1998 Core Spray Sys Insp Program - 17R ML20154L3051998-10-14014 October 1998 Safety Evaluation Accepting Licensee Request to Defer Insp of 79 Welds from One Fuel Cycle at 17R Outage ML20154Q3371998-09-30030 September 1998 Rev 8 to Oyster Creek Cycle 17,COLR ML20154L5571998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Oyster Creek Nuclear Generating Station.With ML20151V6311998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Oyster Creek Nuclear Generating Station.With ML20237D5691998-08-31031 August 1998 Rev 0 to MPR-1957, Design Submittal for Oyster Creek Core Plate Wedge Modification ML20237D5711998-08-18018 August 1998 Rev 0 to SE-000222-002, Core Plate Wedge Installation ML20237B0131998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Oyster Creek Nuclear Generating Station ML20236R0511998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Oyster Creek Nuclear Generating Station ML20249B2981998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Oyster Creek Nuclear Station ML20248F3531998-05-21021 May 1998 Part 21 Rept Re Electronic Equipment Repaired or Reworked by Integrated Resources,Inc from Approx 930101-980501.Caused by 1 Capacitor in Each Unit Being Installed W/Reverse Polarity. Policy of Second Checking All Capacitors Is Being Adopted ML20247F1891998-05-0505 May 1998 Risk Evaluation of Post-LOCA Containment Overpressure Request ML20247G0581998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Oyster Creek Nuclear Generating Station ML20216K0341998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Oyster Creek Nuclear Generating Station ML20151Y4651998-03-31031 March 1998 Non-proprietary Version of Rev 1 to GENE-E21-00143, ECCS Suction Strainer Hydraulic Sizing Rept ML20217A4631998-03-23023 March 1998 Safety Evaluation Accepting Use of Three Heats/Lots of Hot Rolled XM-19 Matl in Core Shroud Repair Assemblies Re Licenses DPR-16 & DPR-59,respectively ML20212E2291998-03-0404 March 1998 Rev 11 to 1000-PLN-7200,01, Gpu Nuclear Operational QAP, Consisting of Revised Pages & Pages for Which Pagination Affected ML20216J0841998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Oyster Creek Nuclear Generating Station ML20203B2781998-02-16016 February 1998 10CFR50.59(b) Rept of Changes to Oyster Creek Systems & Procedures ML20203A3801998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Oyster Creek Nuclear Generation Station ML20198P1791997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Oyster Creek Nuclear Station ML20217C7591997-12-31031 December 1997 1997 Annual Environmental Operating Rept for Oyster Creek Nuclear Generating Station ML20197E9131997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for Oyster Creek Nuclear Station ML20199E4561997-11-13013 November 1997 Safety Evaluation Accepting Ampacity Derating Analysis in Response to NRC RAI Re GL-92-08, Thermo-Lag 330-1 Fire Barriers, for Plant ML20199D4381997-10-31031 October 1997 Monthly Operating Rept for Oct 1997 for Oyster Creek Nuclear Station ML20202E8511997-10-21021 October 1997 Rev 0 to Scenario 47, Gpu Nuclear Oyster Creek Nuclear Generating Station Emergency Preparedness (Nrc/Fema Evaluated) 1997 Biennial Exercise. Pages 49 & 59 of Incoming Submittal Were Not Included ML20211M9481997-10-0303 October 1997 Supplemental Part 21 Rept Re Condition Effected Emergency Svc Water Pumps Supplied by Bw/Ip Intl Inc to Gpu Nuclear, Oyster Creek Nuclear Generation Station.No Other Nuclear Generating Stations Effected by Notification ML20198J7361997-09-30030 September 1997 Monthly Operating Rept for Sept 1997 for Oyster Creek Nuclear Generating Station ML20211B7461997-09-24024 September 1997 Part 21 Rept Re Failure of Emergency Service Water Pump Due to Threaded Flange Attaching Column to Top Series Case Failure.Caused by Dissimilar Metals.Pumps in High Ion Svc Will Be Upgraded to 316 Stainless Steel Matl ML20210V0181997-08-31031 August 1997 Monthly Operating Rept for Aug 1997 for Oyster Creek Nuclear Generating Station ML20210L2961997-07-31031 July 1997 Monthly Operating Rept for Jul 1997 for Oyster Creek Nuclear Station ML20149F9961997-07-18018 July 1997 Safety Evaluation Re Gpu Nuclear Operational Quality Assurance Plan,Rev 10 for Three Mile Island Nuclear Generating Station,Unit 1 & Oyster Creek Nuclear Generating Station ML20196H0111997-07-11011 July 1997 Special Rept 97-001:on 970620,removed High Range Radioactive Noble Gas Effluent Monitor (Stack Ragems) from Service to Allow Secondary Calibr IAW Master Surveillance Schedule. Completed Calibr on 970628 & Returned Stack Ragems to Svc ML20210L3081997-06-30030 June 1997 Corrected Page to MOR for June 1997 for Oyster Creek Nuclear Generating Station ML20141H2051997-06-30030 June 1997 Monthly Operating Rept for June 1997 for Oyster Creek Nuclear Station 1999-09-30
[Table view] |
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RAFETY EVALUATION BY THE OFFICE QF NUCLEAR REACTOR REGULATION I
AMPACITY DERATING ISSUES GENERAL P!,'BLIC UTILITlE8 NUCLEAR CORPQMTION OY8TER CREEK NUCLEAR GENERATING STAIlON DOCKET NO. 80 219 l I i RACKGROUND By letter dated November 25,1996, General Public Utilities (GPU) Nuclear (the licensee) submitted a response to the NRC Request for Additionalinformation (L'.!) ralated to l Generic Letter (GL) 92 08, 'Thermo Lag 3301 Fim Barriers," for the Oyster Creek TNcioar [
Generating Station (OCNGS). ,
The stan RAl dated August 26,1996, had identified a number of open issues and concems i requiring clarification by the licensee. The subject licensee submittal contained the ruponse to stan questions regarding lls empacity methodology. The stan evaluation of the ampacity -
dorating methodology for OCNGS follows, l l i' '
EVALUATION 1 After reviewing the Scentse's submittals and Sa dia National Laboratories (SNL) Technical Letter Report (see Attachment 2), the sta# agrees with the SNL analyses and conclusions. The ampacity dorating analysis questions, the licensee's response, SNL observations and the stefrs evaluation of the responses follow.
Amoscity Deratina Analysis Review
. Question 1 The licentee stated that Thermo Lag was installed on conduits using preformed sections and concludes that this would leave no air gaps. However, the sta# agrees with its contractor that an air gap between the outer surface of the conduit and the inner surface l of the barrier would be expected unless specific steps are taken to eliminate this gap.
This may invalidate the licensee's comparison to the Tennessee Valley Authority (TVA) test results because TVA installation procedures for conduits specifically called for
! eliminating this air gap by "probuttering" the er' tire inner surface of the barrier sections.
The licensee must further assess the applicability of the TVA test results to its own installations recognizing this potential di#erence in installation procedures.
. Enclosure 1 1
M'1E!
'F EsNtv PDR .
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_ ._ _ _ _ _ . _ . - _ _ _ _ . = _ _ _ _ . _ . . _ - _ . _ _
7, 4
Licensee Resoonse t in its submittal dated November 25,1996, the licensee stated that the TVA 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> l
barrier test configurations were not prebuttered on the entire surface of the barrier sections. The Omega Point Laboratory Reports No. 11960-97337 and 97338,
- Electrical Test to Determine the Ampacity Derating of an Electrical Raceway Fire Barrier System for Class IE Electrical Circuits," dated August 21,1995, describe the test barrier material hatallation highlights and specifies that the enclosure for the conduit was constructed from nominal one and one fourth inch thick preformed conduit sections which were dry fit to the conduit and secured with stainless steel bands. The licensee i contends that the presence or lack of an air gap is not applicable because the subject OCNGS Insta!led barriers and the tested TVA 3-hour barriers are comparable in terms of construction.
Staff Resoonse The information provided by the licensee fully resolves the staff's concems.
. Question 2 The licensee needs to provide complete calculations in order to evaluate the appropriateness of its ampacity assessment approach and implementation. Otherwise, specific and complete examples of the ampacity derating calculations illustrating all aspects of those calculations in detail (baseline ampacity with source, cable characteristics, conduit size and type, percent fill, number of conductors, fire barrier rating, etc.) should be provided for typical 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> conduits, typical 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> conduits, and typical air drops. See attached SNL Report for details.
Licensee Resoonse in its submittal dated November 25,1996, the licensee provided one example of the calculations in the following applications: (1) a conduit using a three hour barrier; (2) a conduit using a ene hour barricr; and (3) an air drop barrier.
Staff Resoonse The information provided by the licenses fully resolves the staff's concems.
Questiona ,
it appears that the licensee has assumed an ambient temperature of 30'C for items 15, 16,17, and 31. Ambient temperature of 30'C would not be expected to bound the environmental conditions in most typical plant areas. The licensee should review the assumed ambient temperatures used in the calculation and provide adequate justification for the assumptions.
Licensee Resoonte 2
. ".c .
]
I in its submittel dated November 25,1996, the licensee stated that the OCNGS ambient cable deratings based on 30*C, as previously submitted to the staff have been revised to incorporate more conservative deratings based on an ambient temperature of 40'C.
Staff Resoonse The information provided by the licensee fully resolves the staff's concems.
. Question 4 lt appears that the licensee has assumed an 11% " bounding" ampacity derating factor
( ADF) for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire wraps including air drops. This value fails to bound the Texas Utilities (TU) results in which air drop ADF values in the range of 21.2% to 31.8% were reported. Ampacity assessments for cables involving air drop configurations r,hould be pcformed using realistic estimate of the fire barrier ADF impact.
Licensee Resoonse in its submittal dated November 25,1996, the licensee states that OCNGS Circuits 62-93 and 62100 are enclosed by a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> air drop fire barrier. The subject air drop configurations are each approximately 2 to 3 feet in length. The total length of each circuit is over 500 feet. The licensee contends based on the rationale provided by National Electric Code (NEC) 31015(c) which allows that no additional derating facior need to be applied for circuits which are less than 10% of the totallength of circuit or a maximum of 10 feet that .1o additional derating needs to be applied other than that for the conduit fire barrier (i.e.,11% ADF) since the air drop configuration length is so short Staff Resonagg The Information provided by the licensee adequately addresses the concern regarding a misapplication of Ampacity Correction Factor (ACF) values for air drop fire barriers.
Instead the conduit derating value of 11% is expected to conservatively bound the very short air drop barrier length which should not experience significant localized heating effects. The licensee response fully resolves the staff's concerns.
. Question 5 The calculated load current for battery chargers (items 5 and 6) is shown as 12.5 amps (actual). This current could be much higher if the battery is discharged. Prov!ds a technical basis for the acceptability of the cable during charging the fully discharged battery for the required duration.
Licensee Resoonse in its submittal dated Nosember 25,1996, the licensee cited that battery discharge conditions are only rarely experienced during a two year outage cycle. During such periods current loads are monitcred by plant procedure. The licensee considered a l
l
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~~
e e .
I
, potential 10% overload condition on the maximum current load for the subject battery chargers.
Staff Retoonse The licensee response is adequate to resolve the identified concem. Specifically, the licensee has allowed for a 10% overload condition on the applicable circuit which is by design limitsd to its maximum rated current load in addition, the licensee has cited existing plant procedures to identify and res .ve any overload conditions.
- SNL Observations During the review of the licensee submittal dated November 25,1996, SNL identified the following observations of a minor nature:
- 1) SNL noted that the licensee adjusted all of the NEC tabulated values using an ACF of 0.91 to derate for a plant ambient of 40'C. Ampacity limits for the licensee's 5 kV cables appear to be have taken from NEC Table 310 73, and these table values are already based on 40'C ambient temperature.
Staff Comment: The apparent error results in additional conservatism to the ampacity derating assessment of the subject cables. Therefore, there is no materialimpact on the staff evaluation conclusions.
- 2) "NL notes that the application of the NEC conduit conductor count correction factors was unclear how the set of NEC values that was used in the subject analyses; either the pre 1990 values which are based on 50% los d diversity assumptions or the post 1990 valves which take no credit for loading diversity.
This observation willimpact cladded conduits with 10 or more conductors.
Staff Comment: Giv:n the limited potentialimpact for conduits with 10 or more conductors and the available ampacity margin which should be sufficient to bound a change in the specific ACF value it is anticipated that this concern will have a material impact on the staff evaluation conclusions.
Anolication of Amoacity Derating Methodoloav The licensee's analysis begins by utilizing the baseline ampacity limits for individual cables based on NEC Handbook values. Note that all of the installed Thermo-Lag fire barriers protecting electrical raceways are associated with conduits or air drop configurations only. For each baseline ampacity limit the va'.ae is adjusted for ambient temperature and the number of conductors in the conduits. Finally, the licenses appi:es an ACF value to reflect the impact of the fire barrier system. The final result is ;:n estimate of the fully derated ampacity limit for tha individual cables. Given these estimates the licensee then compares the actualinstalled service loading to the analyzed ampacity limits. The licensee has concluded that the estimated ;
ampacity limits bound the installed service loads at OCNGS.
4
The staff finds that the SNL observations as previously discussed to be items for later onsite review and verification b/ the staff after they have been conveyed to the licensee for disposition. The apparent error involving the unnecessary ambient temperature correction results in a conservative treatment for the applicable 5 kV cables. This item can be addressed at the licensee's dicetetion. The potential concern pertaining to the application of the NEC conduit conductor count correction factors is limited in scope such that given the significant level of margin which has been demonstrated for most cables, it is not anticipated that it will have a materialimpact on the licensee's ampacity derating assessments. Therefore, both items are minor in nature and do not affect the overall staff conclusions regarding the licensee's ampacity derating methodology.
The licensee utilized in its assessment of the fire barrier impact, i.e., ACF values, the results from the Texas Utilibes Electric (TUE) and Tennessee Valley Authority (TVA) ampacity derating tests. Given the licensee is now relying on industry ampacity derating test data (i.e., TUE test data) which has been reviewed and accepted by the staff for its ampacity derating assessment approach the staff finds that the actual operating conditions do not result in exceeding cable ampacity limits for the applicable electihal raceways protected by Thermo Lag fire barriers at OCNGS.
The onsite review of the licensee implementation of its methodology is outside of the scope of this evaluation. The licensee revised calculations are available for onsite review and verification by the staff.
C.QMCLUSIONS ,
From the above evaluation, the staff concludes that no significant safety hazards are introduced through use of the licensee's ampacity derating methodology. Therefore, the application of Thermo-Lag fire barriers to enclose cables at Oyster Creek Nuclear Generating Station does not represent a safety concern with resoect to ampar" 5 9en the licensee revision to its design calculation it is recommended, that the staff evaiuation be used in a follow up site inspection to verify implementation of the licensee changes to its design documentation especially the potential concern regarding the application of the NEC conduit conductor count correction factors (see Section 2.5 of Attachment 2).
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