|
---|
Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20212L1141999-10-0101 October 1999 Safety Evaluation Granting Request for Exemption from Technical Requirements of 10CFR50,App R,Section III.G.2.c ML20212F5261999-09-22022 September 1999 SER Approving Request Reliefs 1-98-001 & 1-98-200,parts 1,2 & 3 for Second 10-year ISI Interval at Arkansas Nuclear One, Unit 1 ML20211F4281999-08-25025 August 1999 Safety Evaluation Concluding That Licensee Provided Acceptable Alternative to Requirements of ASME Code Section XI & That Authorization of Proposed Alternative Would Provide Acceptable Level of Quality & Safety ML20207E7231999-06-0202 June 1999 Safety Evaluation Authorizing Proposed Alternative Exam Methods Proposed in Alternative Exam 99-0-002 to Perform General Visual Exam of Accessible Areas & Detailed Visual Exam of Areas Determined to Be Suspect ML20206M7711999-05-11011 May 1999 SER Accepting Relief Request from ASME Code Section XI Requirements for Plant,Units 1 & 2 ML20206F0691999-04-29029 April 1999 Safety Evaluation Accepting Licensee Re ISI Plan for Third 10-year Interval & Associated Requests for Alternatives for Plant,Unit 1 ML20205M6941999-04-12012 April 1999 Safety Evaluation Granting Relief for Second 10-yr Inservice Inspection Interval for Plant,Unit 1 ML20205D6061999-03-31031 March 1999 Safety Evaluation Supporting Licensee Proposed Approach Acceptable to Perform Future Structural Integrity & Operability Assessments of Carbon Steel ML20205D4711999-03-26026 March 1999 SER Accepting Util Proposed Alternative to Employ Alternative Welding Matls of Code Cases 2142-1 & 2143-1 for Reactor Coolant System to Facilitate Replacement of Steam Generators at Arkansas Nuclear One,Unit 2 ML20204B1861999-03-15015 March 1999 Safety Evaluation Authorizing Licensee Request for Alternative to Augmented Exam of Certain Reactor Vessel Shell Welds,Per Provisions of 10CFR50.55a(g)(6)(ii)(A)(5) ML20198M7841998-12-29029 December 1998 SER Accepting Util Proposal to Use ASME Code Case N-578 as Alternative to ASME Code Section Xi,Table IWX-2500 for Arkansas Nuclear One,Unit 2 ML20196F4911998-12-0101 December 1998 SER Accepting Request for Relief ISI2-09 for Waterford Steam Electric Station,Unit 3 & Arkansas Nuclear One,Unit 2 ML20195C4841998-11-0606 November 1998 SER Accepting QA Program Change to Consolidate Four Existing QA Programs for Arkansas Nuclear One,Grand Gulf Nuclear Station,River Bend Station & Waterford 3 Steam Electric Station Into Single QA Program ML20154J2471998-10-0909 October 1998 SER Accepting Inservice Testing Program,Third ten-year Interval for License DPR-51,Arkansas Nuclear One,Unit 1 ML20248D7491998-05-28028 May 1998 Safety Evaluation Accepting Licensee Request for Relief from ASME Code Repair Requirements for ASME Code Class 3 Piping ML20217A7211998-04-17017 April 1998 Safety Evaluation Supporting Proposed Alternative for ANO-1 to Implement Code Case N-533 (w/4 H Hold Time at Test Conditions Prior to VT-2 Visual Exam) ML20217P8281998-04-0707 April 1998 Safety Evaluation Accepting Relief Authorization for Alternative to Requirements of ASME Section Xi,Subarticle IWA-5250 Bolting Exam for Plants,Per 10CFR50.55a(a)(3)(i) ML20216D6111998-03-12012 March 1998 Safety Evaluation Supporting Amend 188 to License NPF-6 ML20199H3711997-11-19019 November 1997 SER Accepting Approving Request Relief from Requirements of Section XI, Rule for Inservice Insp of NPP Components, of ASME for Current or New 10-year Inservice Insp Interval IAW 50.55(a)(3)(i) of 10CFR50 ML20216E9921997-09-0404 September 1997 Safety Evaluation Accepting 970623 Request for Relief Re Authorization for Use of ASME Code Case N-416-1 & N-532,ISI Program for Listed Plants ML20141H8411997-07-30030 July 1997 Safety Evaluation Accepting Use of Code Case N-508-1 for All Four Plants for Rotation of Serviced Snubbers & Pressure Relief Valves for Purpose of Testing in Lieu of ASME Code ML20138K0561997-05-0505 May 1997 SER Approving Licensees IPE Process Capable of Identifying Severe Accidents & Severe Accident Vulnerabilities,For Plant,Unit 2 ML20149M4221996-12-12012 December 1996 Safety Evaluation Supporting Update Insvc Insp Programs to 1992 & Portions of 1993 ASME Boiler & Pressure Vessel Code, Sect XI for Licenses DPR-51,NPF-6,NPF-38,NPF-29 & NPF-47. Technical Ltr Rept Encl ML20134P8411996-11-25025 November 1996 Safety Evaluation Denying Request for Relief 96-001 Re Second 10-yr Interval ISI Program Plan,Due to Failure to Provide Basis for Impracticality ML20107F5611996-04-17017 April 1996 Safety Evaluation Providing Guidance on Submitting plant- Specific Info W/Respect to IST Program Alternatives Request ML20149K9451996-02-16016 February 1996 Safety Evaluation Authorizing Relief Request for Second 10-yr Interval IST Program Plan for Pumps & Valves at Facility ML20058L6111993-12-13013 December 1993 Safety Evaluation Approving Second ten-year Interval Inservice Insp Request for Relief Re Use of IWA-5250 Requirements Listed in 1992 Edition of ASME Code ML20058F6661993-11-24024 November 1993 Safety Evaluation Accepting Licensee Proposed Use of New DG as Alternate AC Power Source for Coping W/Sbo Subject ML20056H1331993-08-23023 August 1993 Supplemental Safety Evaluation Re Conformance to Reg Guide 1.97.Plant Design in Conformance W/Guidance of Subj Reg Guide ML20056H0001993-08-19019 August 1993 Safety Evaluation Accepting Licensee 920918 Response to GL 87-02,suppl 1 ML20128C9181993-01-22022 January 1993 Safety Evaluation Supporting Inservice Testing Program Relief Requests for Pumps & Valves ML20126H8661992-12-30030 December 1992 Safety Evaluation Granting Relief from Certain Inservice Insp Requirements of Section XI of ASME Boiler & Pressure Vessel Code,Determined to Be Impractical to Perform ML20126F7571992-12-18018 December 1992 Safety Evaluation Accepting Util Conceptual Design for Proposed Alternate Ac Power Source ML20062A5751990-10-10010 October 1990 Safety Evaluation Re Station Blackout Rule.Util Response Does Not Conform W/Station Blackout Rule ML20062A5881990-10-10010 October 1990 Safety Evaluation Re Station Blackout Rule.Util Response Does Not Conform W/Station Blackout Rule ML20059A7081990-08-17017 August 1990 Sser Concluding That Rochester Instrument Sys Model SC-1302 Isolation Device Acceptable for Use at Plant for Interfacing SPDS W/Class IE Circuits ML20056A7511990-08-0707 August 1990 Safety Evaluation Accepting Licensee Fire Barrier Penetration Seal Program & Commitment to Complete 100% Review of All Tech Spec Fire Penetration Seals by 911231 ML20062C8341990-05-24024 May 1990 Safety Evaluation Granting Relief from Certain Inservice Insp Requirements of ASME Code,Section Xi,Per 881103 & 890823 Requests ML20245K3671989-08-11011 August 1989 Safety Evaluation Accepting Licensee Actions in Response to 890120 High Pressure Injection Backflow Event ML20247N7451989-07-31031 July 1989 Safety Evaluation Concluding That Isolation Devices Acceptable for Use in Spds,Contingent on Licensee Submittal of Followup Evaluation Verifying That Failure of RIC SC-1302 Was Randomly Deficient Device Prior to Testing ML20247A8371989-07-11011 July 1989 Safety Evaluation Re Generic Ltr 83-28,Item 4.5.3 Concerning Reactor Trip Sys Reliability for All Domestic Operating Reactors ML20246J5421989-07-0707 July 1989 Safety Evaluation Re NRC Audit of Util Resolution of IE Bulletin 79-27.IE Bulletin Concerns Adequately Resolved for Facility.Periodic Test Program for Devices Recommended to Be Developed by Licensee ML20245K5971989-06-21021 June 1989 Safety Evaluation Concluding That Diverse Scram Sys & Diverse Initiation of Turbine Trip Meet Requirements of ATWS Rule (10CFR50.62) ML20245F3921989-04-25025 April 1989 Safety Evaluation Granting Util Relief from ASME Section XI Insp Requirements for Reactor Coolant Pump Casing Weld Indications Due to Impractical Requirements,Per Util 881027 Request & 10CFR50.55a ML20247E0191989-03-23023 March 1989 Safety Evaluation Supporting Instrumentation for Detection of Inadequate Core Cooling for Plants ML20206F5021988-11-15015 November 1988 Safety Evaluation Supporting Transfer of Operating Responsibility to Sys Energy Resources,Inc ML20205H8751988-10-25025 October 1988 Safety Evaluation Supporting Util 861124 & 880603 Responses to Generic Ltr 86-06,TMI Action Item II.K.3.5 Re Automatic Trip of Reactor Coolant Pumps ML20151D4131988-07-12012 July 1988 Safety Evaluation Supporting Util 831105 Responses to Generic Ltr 83-28,Item 4.5.2 Re Reactor Trip Sys Reliability Online Testing ML20236A7581987-10-15015 October 1987 Evaluation Supporting Justification for Continued Operation Re High Reactor Bldg Temps ML20235W7011987-07-15015 July 1987 Safety Evaluation Re HPI Makeup Nozzle Cracking.Util Agreement to Record HPI Flowrate & Duration of Flow During HPI Actuation When SPDS Available Acceptable 1999-09-22
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217L8931999-10-31031 October 1999 Rev 1 to BAW-10235, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs ML20212L1141999-10-0101 October 1999 Safety Evaluation Granting Request for Exemption from Technical Requirements of 10CFR50,App R,Section III.G.2.c 0CAN109902, Monthly Operating Repts for Sept 1999 for Arkansas Nuclear One,Units 1 & 2.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Arkansas Nuclear One,Units 1 & 2.With ML20216J6271999-09-27027 September 1999 Rev 0 to CALC-98-R-1020-04, ANO-1 Cycle 16 Colr ML20212F5261999-09-22022 September 1999 SER Approving Request Reliefs 1-98-001 & 1-98-200,parts 1,2 & 3 for Second 10-year ISI Interval at Arkansas Nuclear One, Unit 1 0CAN099907, Monthly Operating Repts for Aug 1999 for Ano,Units 1 & 2. with1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Ano,Units 1 & 2. with ML20211F4281999-08-25025 August 1999 Safety Evaluation Concluding That Licensee Provided Acceptable Alternative to Requirements of ASME Code Section XI & That Authorization of Proposed Alternative Would Provide Acceptable Level of Quality & Safety 0CAN089904, Monthly Operating Repts for July 1999 for Ano,Units 1 & 2. with1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Ano,Units 1 & 2. with ML20210K8831999-07-29029 July 1999 Non-proprietary Addendum B to BAW-2346P,Rev 0 Re ANO-1 Specific MSLB Leak Rates 0CAN079903, Monthly Operating Repts for June 1999 for Ano,Units 1 & 2. with1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Ano,Units 1 & 2. with ML20207E7231999-06-0202 June 1999 Safety Evaluation Authorizing Proposed Alternative Exam Methods Proposed in Alternative Exam 99-0-002 to Perform General Visual Exam of Accessible Areas & Detailed Visual Exam of Areas Determined to Be Suspect ML20196A0191999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Arkansas Nuclear One,Units 1 & 2.With ML20196A6251999-05-31031 May 1999 Non-proprietary Rev 0 to TR BAW-10235, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs ML20195D1991999-05-28028 May 1999 Probabilistic Operational Assessment of ANO-2 SG Tubing for Cycle 14 ML20206M7711999-05-11011 May 1999 SER Accepting Relief Request from ASME Code Section XI Requirements for Plant,Units 1 & 2 0CAN059903, Monthly Operating Repts for Apr 1999 for Ano,Units 1 & 2. with1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Ano,Units 1 & 2. with ML20206F0691999-04-29029 April 1999 Safety Evaluation Accepting Licensee Re ISI Plan for Third 10-year Interval & Associated Requests for Alternatives for Plant,Unit 1 ML20205M6941999-04-12012 April 1999 Safety Evaluation Granting Relief for Second 10-yr Inservice Inspection Interval for Plant,Unit 1 ML20205D6061999-03-31031 March 1999 Safety Evaluation Supporting Licensee Proposed Approach Acceptable to Perform Future Structural Integrity & Operability Assessments of Carbon Steel ML20205R6351999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Ano,Units 1 & 2. with ML20205D4711999-03-26026 March 1999 SER Accepting Util Proposed Alternative to Employ Alternative Welding Matls of Code Cases 2142-1 & 2143-1 for Reactor Coolant System to Facilitate Replacement of Steam Generators at Arkansas Nuclear One,Unit 2 ML20204B1861999-03-15015 March 1999 Safety Evaluation Authorizing Licensee Request for Alternative to Augmented Exam of Certain Reactor Vessel Shell Welds,Per Provisions of 10CFR50.55a(g)(6)(ii)(A)(5) 0CAN039904, Monthly Operating Repts for Feb 1999 for Ano,Units 1 & 2. with1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Ano,Units 1 & 2. with ML20212G6381999-02-25025 February 1999 Ano,Unit 2 10CFR50.59 Rept for 980411-990225 ML20203E4891999-02-11011 February 1999 Rev 1 to 97-R-2018-03, ANO-2,COLR for Cycle 14 ML20199F0351998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Ano,Units 1 & 2 ML20198M7841998-12-29029 December 1998 SER Accepting Util Proposal to Use ASME Code Case N-578 as Alternative to ASME Code Section Xi,Table IWX-2500 for Arkansas Nuclear One,Unit 2 0CAN129805, LER 98-S02-00:on 981124,security Officer Found Not to Have Had Control of Weapon for Period of Approx 3 Minutes Due to Inadequate self-checking to Ensure That Weapon Remained Secure.All Security Officers Briefed.With1998-12-11011 December 1998 LER 98-S02-00:on 981124,security Officer Found Not to Have Had Control of Weapon for Period of Approx 3 Minutes Due to Inadequate self-checking to Ensure That Weapon Remained Secure.All Security Officers Briefed.With ML20196F4911998-12-0101 December 1998 SER Accepting Request for Relief ISI2-09 for Waterford Steam Electric Station,Unit 3 & Arkansas Nuclear One,Unit 2 ML20198D2441998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Ano,Units 1 & 2. with ML20199F7401998-11-16016 November 1998 Rev 9 to ANO-1 Simulator Operability Test,Year 9 (First Cycle) ML20195B4801998-11-0707 November 1998 Rev 20 to ANO QA Manual Operations ML20195C4841998-11-0606 November 1998 SER Accepting QA Program Change to Consolidate Four Existing QA Programs for Arkansas Nuclear One,Grand Gulf Nuclear Station,River Bend Station & Waterford 3 Steam Electric Station Into Single QA Program 0CAN119808, Monthly Operating Repts for Oct 1998 for Ano,Units 1 & 2. with1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Ano,Units 1 & 2. with ML20197H0741998-10-29029 October 1998 Rev 1 to Third Interval ISI Program for ANO-1 ML20155C1351998-10-26026 October 1998 Rev B to Entergy QA Program Manual ML17335A7641998-10-22022 October 1998 LER 98-004-00:on 980923,inadvertent Actuation of Efs Occurred During Surveillance Testing.Caused by Personnel Error.Personnel Involved with Event Were Counseled & Procedure Changes Were Implemented.With 981022 Ltr ML20154J2471998-10-0909 October 1998 SER Accepting Inservice Testing Program,Third ten-year Interval for License DPR-51,Arkansas Nuclear One,Unit 1 0CAN109806, Monthly Operating Repts for Sept 1998 for ANO Units 1 & 2. with1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for ANO Units 1 & 2. with ML20154E2171998-09-28028 September 1998 Follow-up Part 21 Rept Re Defect with 1200AC & 1200BC Recorders Built Under Westronics 10CFR50 App B Program. Westronics Has Notified Bvps,Ano & RBS & Is Currently Making Arrangements to Implement Design Mods 0CAN099803, Monthly Operating Repts for Aug 1998 for ANO Units 1 & 2. with1998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for ANO Units 1 & 2. with ML20237B7671998-08-19019 August 1998 ANO REX-98 Exercise for 980819 ML18066A2771998-08-13013 August 1998 Part 21 Rept Re Deficiency in CE Current Screening Methodology for Determining Limiting Fuel Assembly for Detailed PWR thermal-hydraulic Sa.Evaluations Were Performed for Affected Plants to Determine Effect of Deficiency ML20236X2351998-08-0505 August 1998 Part 21 Rept Re Defect Associated W/Westronics 1200AC & 1200BC Recorders Built Under Westronics 10CFR50,App B Program.Beaver Valley,Arkansas Nuclear One & River Bend Station Notified.Design Mod Is Being Developed 0CAN089804, Monthly Operating Repts for July 1998 for Ano,Units 1 & 21998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Ano,Units 1 & 2 ML20196C7831998-07-30030 July 1998 Summary Rept of Results for ASME Class 1 & 2 Pressure Retaining Components & Support for ANO-1 ML20155H7161998-07-15015 July 1998 Rev 1 to 96-R-2030-02, Revised Reactor Vessel Fluence Determination ML20236R0531998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Ano,Units 1 & 2 ML20249B7791998-06-22022 June 1998 Part 21 Rept Re Findings,Resolutions & Conclusions Re Failure of Safety Related Siemens 4KV,350 MVA,1200 a Circuit Breakers to Latch Closed ML20249B5091998-06-15015 June 1998 SG ISI Results for Fourteenth Refueling Outage 1999-09-30
[Table view] |
Text
f~
l
, , . , purco e* 4 UNITED STATES f j
~
- NUCLEAR REGULATORY COMMISSION A" WASHINGTON, D.C. 20555-0001 l
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO APPENDIX R EXEMPTION REQUEST FOR ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT 2 DOCKET NO. 50-368
1.0 INTRODUCTION
Appendix R," Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1,1979, " to Title 10 of the Code of FederalRegulations (10 CFR), Part 50, establishes fire protection features required to satisfy General Design Criterion 3,' Fire protection," of Appendix A to 10 CFR Part 50 with respect to certain generic issues for nuclear power plants licensed to operate prior to January 1,1979. By letter dated March 22,1983, the staff issued
~
an exemption from the technical requirements of Section Ill.G.2 of Appendix R to the extent that it requires an automatic fire suppression system for the area below the 354-foot elevation of the Arkansas Nuclear One, Unit 2 (ANO-2), intake structure. By letter dated October 8,1997, as supplemented by letter dated February 25,1999, Entergy Operations, Inc. (the licensee), for ANO-2, submitted a revised exemption request for this area.
2.0 EXEMPTION REQUESTED The licensee requested an exemption from the technical requirements of Section Ill.G.2 of Appendix R to the extent that it requires an automatic fire suppression system and an automatic fire detection system for the area below the 354 foot elevation of the ANO-2 intake structure.
3.0 DISCUSSION The ANO-2 intake structure is about 32 feet by 26 feet on three levels. There are no rated fire barriers between the three levels. Below the 354-foot elevation there are three intake bays, which contain service water (SW) piping and conduits. The bays are about 7 feet by 32 feet and are separated from one another by 2-foot thick, non-rated concrete walls. The bays are separated from ground level by an 18-inch thick, non-rated concrete slab on metal decking.
The intake bays are typically flooded with water to a depth of about 16 feet. The water is normally provided through a sluice gate. The ceiling height is about 30 feet above the bottom of the bay and about 14 feet above the normal pool level. Access to the intake bays is restricted and is obtained through hatches at the 354 foot elevation.
Redundant post-fire safe shutdown equipment on all levels of the intake structure is limited to equipment associated with the SW system. SW is required to be available to supply cooling water for various safe shutdown components including the diesel generators and the shutdown cooling heat exchangers. In addition, SW can be aligned to the emergency feedwater system if the desired condensate sources become depleted. One flow path is needed to support 9910070162 991001 PDR ADOCK 05000368 l P PM l
L __
n L
1 -
N 2 post-fire safe shutdown. The SW system consists of two independent seismic category I flow paths, which furnish water to two independent trains of 100 percent capacity emergency safety feature equipment, and two nonseismic category I flow paths. The SW system is provided with two dedicated SW pumps (2P4A and 2P4C) and a swing SW pump (2P48) that can be aligned to either of the two flow paths. Each pump provides 100 percent of the required flow for the ,
respective flow path. During plant operations (Modes 1 through 5), the ANO-2 technical specification requires that two SW trains be operable. The possible SW pump alignments are ]
j SW pumps 2P4A and 2P48, SW pumps 2P4A and 2P4C, or SW pumps 2P4C and 2P4B. The power cable arrangements are as follows: conduit EA 1007 contains the red train power supply cable to SW pump 2P4A; conduit EA2036 contains the green train power supply cable to swing !
SW pump 2P4B; and conduit EA2007 contains the green train power supply cable for SW pump 2P4C. Conduits EA1007 and EA2036 are protected by separate 1-hour fire-rated Hemyc
]
fire barriers. Below the 354-foot elevation, these conduits are also encapsulated in a common -
galvanized sheet metal moisture barrier. Conduit EA2007, which is located about 6 feet from the moisture barrier containing conduits EA1007 and EA2036, is covered by a Thermo-Lag barrier. The licensee stated that it does not take credit for the Thermo-Lag barrier to meet the requirements of Appendix R. Conduit EA1008, which contains the red train power supply to swing SW pump 2P48, is embedded in the concrete slab at the elevation of 354 feet and does not enter the bay.
SW bay "A" contains redundant cables required to support post-fire safe shutdown. The licensee stated that the 2P4C/2P48 SW pump combination with SW pump 2P4B aligned to the red train power is the only pump alignment that would be utilized during normal operations in Modes 1 through 5 with SW bay "A" isolated and drained. During the recovery from a fire, the time critical function is to supply cooling water to the diesel generators. The licensee stated that, on the basis of its calculations, the diesel generators (and therefore the SW system components) are not required to be operated during the first 30 minutes of a fire event. The licensee allows the operators to manually align the SW system because sufficient time is available to complete the alignment.
Power and control cables for the sluice gates, which are routed in conduits EB1008, EC1140, EB2008, and EC2132, are also located in the SW intake bays. Sluice gate valves 2CV1470-1, 2CV1472-5, and 2CV1474-2 are normally open, which corresponds to the safe shutdown position. The redundant control cables are separated horizontally by about 8 feet. As stated previously, the time critical function of the SW system is to provide cooling to the diesel generators. The licensee stated that if a fire were to cause the sluice gates to spuriously close, adequate time would be available before service water was required to manually realign any affected component.
The in situ combustible loading below elevation 354 feet of the intake structure is limited to PVC pip!ng and Thermo-Lag fire barrier material. There are no cable trays or exposed cables in the SW bays. The licensee administratively controls access to the bays as confined spaces and foreign material exclusion areas, thus limiting access by personnel durng routine operations and reducing the likelihood that uncontrolled transient combustibles will be introduced into or accumulate in the bays. In addition, the licensee's administrative procedures limit the transient combustibles to 5 pounds unless personnel are conthuously present in the area. In such cases, the personnel could be either the craft personnel responsible for using the combustible
! materials or a continuous fire watch.
r The two upper levels (elevations 354 feet and 366 feet) of the intake structure have flame detectors or ionization smoke detectors and a preaction fire suppression system. Fire protection in SW bay "A" consists of the 1-hour fire-rated Hemyc barriers for conduits EA1007 l and EA2036. Portable fire extinguishers and a fire hose station are available for manual fire fighting.
4.0 EVALUATION l l
i The area below the 354-foot elevation of the ANO-2 intake structure does not meet the technical requirements of Section Ill.G.2.c of Appendix R to 10 CFR Part 50 because, although a 1-hour-rated fire barrier is provided as discussed below, fire detectors and an automatic fire l suppression system are not provided. l The in-situ combustibles in SW bay "A" and the administratively allowed quantity of transient l combustibles (5 pounds) do not pcse a credible fire threat to the SW pump cables. In the staff's view, a fire involving transient combustibles in excess of the administratively allowed i quantity is the only type of fire that could damage redundant SW pump power cables. The l licensee has addressed this threat by protecting both the red train power supply cable for GW pump 2P4A and the green train power supply cable for swing SW pump 2P4B with 1-hour i fire rated barriers, by embedding the red train power supp;y cable for SW swing pump 2P4B in concrete which provides an equivalent 1-hour-rated fire barrier, and by administratively requiring the presence of craft personnel or a fire watch, if the administrative transient combustible limit is exceeded.
A fire involving transient combustibles could be extinguished by the craft personnel or the fire watch during its incipient stage, in the event the fire grows beyond the incipient stage before it is extinguished, the craft personnel or the fire watch could summon the plant fire brigade. In addition, the smoke and hot gases would be directed upwards into the higher elevations of the intake structure, which are equipped with an automatic fire detection system. Therefore,in the event that a fire in the intake bay is not discovered by the craft personral or the fire watch, it would be detected by the automatic fire detection system and the plant t;e brigade would be dispatched. If the fire exposes the redundant conduits, the 1-hour fire-rated barriers and the concrete embedding would provide fire resistive protection, with margin, for the expected fire hazards and, therefore, provide reasonable assurance that the power cables would not be damaged before the fire either burns itself out or is extinguished by the craft personnel, the fire watch, or the fire brigade. On this basis, the staff concludes that the existing fire protection design features, coupled with the administrative controls, provide reasonable assurance that a fire in SW bay "A" would not damage the redundant SW pump power cables and, therefore, would not adversely affect the ability to achieve and maintain post-fire safe shutdown. The staff also concludes that the installation of fire detectors and an automatic fire suppression system in the area below the 354-foot elevation of the ANO-2.nta. ./.. cu:c ',ou;d not result in a significant increase in the level of fire safety for the redundant SW pumps.
l
- f.
5.0 CONCLUSION
The underlying purpose of Section lil.G.2 of Appendix R is to provide reasonable assurance that at least'one means of achieving and maintaining safe shutdown conditions will remain
. available during and after any postulated fire in the plant. On the basis of its review ano
! evaluation, the staff concludes that fire detectors and an automatic fire suppression system are not needed to satisfy the underlying purpose of Section Ill.G.2.c of Appendix R to 10 CFR Part 50 for the area below the 354-foot elevation of the ANO-2 intake structure, and that there I
would be no undue risk to public health and safety. The licensee's request for an exemption from the technical requirements of Section Ill G.2.c of Appendix R for this area should, therefore, be granted.
l L Principal Contributor: J. A. Holmes Date: October 1, 1999 l
i-