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Category:NRC TECHNICAL REPORT
MONTHYEARML20056E5171993-08-31031 August 1993 Technical Review Rept, Tardy Licensee Actions ML20248F0001989-09-29029 September 1989 Debris in Containment Recirculation Sumps, Technical Review Rept ML20248C1311989-07-17017 July 1989 Diagnostic Evaluation Team Rept for Brunswick Steam Electric Plant,Units 1 & 2 ML20245B6061988-08-31031 August 1988 Inadequate NPSH in HPSI Sys in Pwrs, Engineering Evaluation Rept ML20204J6141988-08-31031 August 1988 AEOD/E807, Pump Damage Due to Low Flow Cavitation ML20196G5251988-06-15015 June 1988 Technical Review Rept T809, Blocked Thimble Tubes/Stuck Incore Detector ML20245H9601988-04-15015 April 1988 BWR Overfill Events Resulting in Steam Line Flooding, AEOD Engineering Evaluation Rept ML20148D0671988-03-17017 March 1988 Headquarters Daily Rept for 880317 ML20148B3291988-03-14014 March 1988 Headquarters Daily Rept for 880314 ML20196H6351988-03-0808 March 1988 Headquarters Daily Rept for 880308 ML20196G8881988-03-0303 March 1988 Headquarters Daily Rept for 880303 ML20147E3961988-01-0606 January 1988 Rept of Interview W/Rg Lagrange on 841206 & 14 to Discuss Info Contained in B Hayes 841017 Memo Identifying Series of Submittals Received from Util Between 1980 & 1984 ML20147E3211988-01-0606 January 1988 Rept of Interview W/Rg Lagrange to Discuss Gpu 830520 & s Re Environ Qualification equipment.Marked-up 850409 Statement from H Hukill Also Encl ML20237L3001987-08-24024 August 1987 AEOD/E709 Engineering Evaluation Rept Re Auxiliary Feedwater Trips Caused by Low Suction Pressure.Draft Info Notice Encl ML20235C9311987-06-23023 June 1987 Rept to ACRS Re Humboldt Bay Unit 3 - Core II ML20212F6581986-12-31031 December 1986 Technical Review Rept, Degradation of Safety Sys Due to Component Misalignment &/Or Mispositioned Control/Selector Switches ML20212D9091986-12-23023 December 1986 Localized Rod Cluster Control Assembly (Rcca) Wear at PWR Plants, Engineering Evaluation Rept ML20212B0321986-12-17017 December 1986 Emergency Diesel Generator Component Failures Due to Vibration, Engineering Evaluation Rept ML20214R4851986-10-0909 October 1986 Initial OL Review Rept for Seabrook Station Unit 1 ML20212K6641986-08-0707 August 1986 Inadvertent Recirculation Actuation Signals at C-E Plants, Technical Review Rept ML20206H0871986-03-0303 March 1986 Allegation Review Data Sheet for Case 4-85-A-013 Re Const Activities.Addl Info Requested from Alleger.Case Closed Due to Lack of Response.Related Info Encl ML20206H0761986-01-21021 January 1986 Allegation Review Data Sheet for Case 4-84-A-085 Re Alteration of Personnel Records.Based on Resolution of Allegation 4-84-A-094,case Closed ML20137X6151986-01-0909 January 1986 Engineering Evaluation of Deficient Operator Actions Following Dual Function Valve Failures ML20234F5601985-12-17017 December 1985 Draft Hazards Analysis ML20234F4751985-12-17017 December 1985 Licensing of Power Reactors by Aec ML20137Y7041985-12-0505 December 1985 AEOD/T515, RHR Svc Water Booster Pump Air Binding at Brunswick Unit 1, Technical Review Rept.Licensee Will Change RHR Svc Water Vent Line During Next Plant Shutdown ML20214T2211985-11-25025 November 1985 Initial OL Review Rept:Millstone Point Unit 3 ML20206H0621985-10-15015 October 1985 Allegation Review Data Sheet for Case 4-85-A-045 Re Inadequate Handling/Installation Procedures for Equipment, Vendor Control Programs & Spare Parts.Based on Insp Rept 50-482/85-22,case Closed ML20206H0371985-10-0202 October 1985 Allegation Review Data Sheet for Case 4-85-A-044 Re Lack of Effective QA Programs & QC Insps.Based on Insp Rept 50-482/85-22,allegation Closed IR 05000482/19850191985-09-30030 September 1985 Allegation Review Data Sheet for Case 4-85-A-050 Re Mishandling of Document Control Program.Concerns Addressed in Insp Rept 50-482/85-19.Dept of Labor & Allegation Cases Closed ML20206H0131985-09-27027 September 1985 Allegation Review Data Sheet for Case 4-84-A-076 Re Vague Administrative Procedures,Calibr Program Not Working,Test Engineer Authority & Harassment.Based on Insp Rept 50-482/85-03,case Closed ML20137B1231985-09-16016 September 1985 HPCS Sys Relief Valve Failures, Engineering Evaluation Rept ML20206G8431985-09-0303 September 1985 Allegation Review Data Sheet for Case 4-84-A-013 Re Improper Termination of Employee Due to Refusal to Weld Laminated Pipe.Welding non-safety Related.Case Closed on 850827.W/ 840315 Telcon Record & Addl Info ML20209G5761985-08-29029 August 1985 AEOD/T509, Inadequate Surveillance Testing Procedures for Degraded Voltage & Undervoltage Relays Associated W/4,160- Volt Emergency Buses, Technical Review Rept.Further AEOD Action Required If Addl Events Identified IR 05000482/19850311985-08-28028 August 1985 Allegation Review Data Sheet for Case 4-85-A-077 Re 6 Rem Exposure in Containment Bldg Due to Pipe Break.Allegation Investigated During Insp 50-482/85-31 on 850715-19 & Found Unsubstantiated ML20206G8051985-08-27027 August 1985 Allegation Review Data Sheet for Case 4-84-A-114 Re Drugs Planted at Plant.Evidence Destroyed in Testing.Based on Insp Rept 50-482/85-03 & Mullikin 850429 Memo,Case Closed ML20206G7781985-08-27027 August 1985 Allegation Review Data Sheet for Case 4-84-A-195 Re Quality First.Fuel Load Issue Resolved in Insp Rept 50-482/85-10. Technical Issues to Be Resolved Prior to Full Power Licensing.Case Closed w/850815 Memo to File ML20209G3051985-08-0909 August 1985 Closure of ECCS Min Flow Valves, Engineering Evaluation Rept.Recommends IE Issue Info Notice to Remind Licensees of Importance of Min Flow Bypass Capability as Essential Pump Protection Feature ML20206H1021985-07-30030 July 1985 Allegation Review Data Sheet for Case 4-84-A-008 Re Improper Const Practices.Insp Rept 50-482/84-12 Issued on 841012 & Closeout Ltr Sent on 850405 ML20206H0801985-07-30030 July 1985 Allegation Review Data Sheet for Case 4-84-A-007 Re Intimidation of QC Inspector.Forwards Documents Closing Allegation.W/O Encls ML20147E4401985-06-20020 June 1985 Rept of Interview W/Cw Smyth on 850510.Smyth Advised of Unfamiliarity W/Environ Qualification Program in Technical Sense & W/Documentation Needed to Qualify Individual components.Marked-up Lw Harding Statement Encl ML20129G3031985-05-13013 May 1985 Valve Stem Susceptibility to IGSCC Due to Improper Heat Treatment, Engineering Evaluation Rept ML20199G0701985-05-0303 May 1985 Partially Withheld Statement of Decision Re Allegation AQ-38 Concerning Alleged Harassment of QC Inspectors Upon Observation of Weld Defects on vendor-inspected Restraints. Allegation Substantiated.Addl Allegation Repts Encl ML20147H0101985-04-16016 April 1985 Draft Summary Rept for Regional Evaluation of Texas Utils Electric Co,Comanche Peak Steam Electric Station ML20206G9151985-03-12012 March 1985 Allegation Review Data Sheet for Case 4-84-A-015 Re Harassment of Mechanical/Welding QC Inspector for Writing Nonconformance Rept Re Improper Welding Amperage by Superintendent.Util Rept Issued & Case Closed ML20147G9901985-01-31031 January 1985 Summary Rept for Regional Evaluation of Texas Utils Electric Co,Comanche Peak Steam Electric Station ML20205Q7691985-01-18018 January 1985 Status Rept Mechanical/Piping Area. Related Info Encl ML20206G8861985-01-0909 January 1985 Allegation Review Data Sheet for Case 4-85-A-004 Re Electrical Installations.Insp Required.Related Info Encl ML20214R5681984-12-31031 December 1984 Shoreham Nuclear Power Station Initial OL Readiness Assessment ML20214T7251984-11-30030 November 1984 Summary Rept for Regional Evaluation of Diablo Canyon Unit 2 1993-08-31
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217N3271999-10-21021 October 1999 Part 21 Rept Re non-linear Oxygen Readings with Two (2) Model 225 CMA-X Containment Monitoring Sys at Bsep.Caused by High Gain Produced by 10K Resistor Across Second Stage Amplifier.Engineering Drawings Will Be Revised BSEP-99-0168, Monthly Operating Repts for Sept 1999 for Bsep,Units 1 & 2. with1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Bsep,Units 1 & 2. with ML20212D0431999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Brunswick Steam Electric Plant,Units 1 & 2 ML20210P9441999-08-10010 August 1999 Safety Evaluation Accepting Licensee Assessment of Impact on Operation of Plant,Unit 1,with Crack Indications of 2.11, 6.36 & 1.74 Inches in Three Separate Jet Pump Risers ML20210P9181999-08-10010 August 1999 Safety Evaluation Authorizing Request for Reliefs CIP-01,02, 06,07,08,09,10 & 11 (with Certain Exceptions) & 12-18,for Second 10-year ISI Interval.Request CIP-04 & 05 Would Result in hardship,CIP-03 Not Required & CIP-11 Denied in Part ML20210N2341999-08-0505 August 1999 SER Accepting Response to NRC GL 87-02, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Unresolved Safety Issues (USI) A-46 ML20210R1191999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Bsep,Units 1 & 2 ML20210R1311999-06-30030 June 1999 Revised Monthly Operating Rept for June 1999 for Bsep,Unit 2 BSEP-99-0118, Monthly Operating Repts for June 1999 for Bsep,Units 1 & 2. with1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Bsep,Units 1 & 2. with BSEP-99-0095, Monthly Operating Repts for May 1999 for Brunswick Steam Electric Plant,Units 1 & 2.With1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Brunswick Steam Electric Plant,Units 1 & 2.With ML20210M8581999-05-14014 May 1999 B214R1 RPV Hydrotest Bolted Connection Corrective Action Evaluation, Rev 0 ML20211L3711999-05-10010 May 1999 Rev 0 to ESR 98-00333, Unit 2 Invessel Feedwater Sparger Evaluation ML20206G1871999-05-0404 May 1999 Safety Evaluation Approving Third 10-year ISI Program Requests for Relief (RR) RR-08,RR-15 & RR-17 BSEP-99-0075, Monthly Operating Repts for Apr 1999 for Brunswick Steam Electric Plant,Unit 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Brunswick Steam Electric Plant,Unit 1 & 2.With ML20206N1791999-04-23023 April 1999 Rev 0 to 2B21-0554, Brunswick Unit 2,Cycle 14 Colr BSEP-99-0059, Monthly Operating Repts for Mar 1999 for Brunswick Steam Electric Plant,Units 1 & 2.With1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Brunswick Steam Electric Plant,Units 1 & 2.With ML20205F9031999-03-30030 March 1999 Safety Evaluation Supporting Proposed Rev to BSEP RERP to Licenses DPR-62 & DPR-71,respectively BSEP-99-0043, Monthly Operating Repts for Feb 1999 for Brunswick Steam Electric Plant,Units 1 & 2.With1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Brunswick Steam Electric Plant,Units 1 & 2.With ML20206N1831999-02-28028 February 1999 Rev 0 to Suppl Reload Licensing Rept for Bsep,Unit 2 Reload 13 Cycle 14 ML20203D7061999-02-0909 February 1999 SER Accepting Proposed Alternatives Contained in Relief Requests PRR-04,VRR-04,VRR-13,PRR-01,PRR-03,VRR-01.VRR-07, VRR-08 & VRR-09 Denied BSEP-99-0005, Monthly Operating Repts for Dec 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With BSEP-98-0231, Monthly Operating Repts for Nov 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With BSEP-98-0218, Monthly Operating Repts for Oct 1998 for Bsep,Units 1 & 2. with1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Bsep,Units 1 & 2. with BSEP-98-0210, Special Rept:On 980824,temp Element 2-CAC-TE-1258-22 Failed. Cause of Failed Temp Element Cannot Be Conclusively Determined.Temp Element Will Be Replaced & Cable Connections Repaired1998-10-30030 October 1998 Special Rept:On 980824,temp Element 2-CAC-TE-1258-22 Failed. Cause of Failed Temp Element Cannot Be Conclusively Determined.Temp Element Will Be Replaced & Cable Connections Repaired ML20154P8151998-10-16016 October 1998 SER Accepting Revised Safety Analysis of Operational Transient of 920117,for Plant,Unit 1 ML20154P8591998-10-16016 October 1998 SER Accepting Equivalent Margins Analysis for N-16A/B Instrument Nozzles for Plant,Units 1 & 2 BSEP-98-0202, Monthly Operating Repts for Sept 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With ML20151Y6211998-09-14014 September 1998 BSEP Rept Describing Changes,Tests & Experiments, for Bsep,Units 1 & 2 ML20151Y6371998-09-14014 September 1998 Changes to QA Program, for Bsep,Units 1 & 2 BSEP-98-0185, Monthly Operating Repts for Aug 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With1998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With ML20151T5021998-08-0505 August 1998 Project Implementation Plan, Ngg Yr 2000 Readiness Program, Rev 2 BSEP-98-0164, Monthly Operating Repts for July 1998 for BSEP Units 1 & 21998-07-31031 July 1998 Monthly Operating Repts for July 1998 for BSEP Units 1 & 2 ML20236T1921998-07-0101 July 1998 Rev 1 to 1B21-0537, Brunswick Unit 1,Cycle 12 Colr ML20236T1961998-07-0101 July 1998 Rev 1 to 2B21-0088, Brunswick Unit 2,Cycle 13 Colr BSEP-98-0142, Monthly Operating Repts for June 1998 for BSEP Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for BSEP Units 1 & 2 ML20236T1971998-06-30030 June 1998 Rev 2 to 24A5412, Supplemental Reload Licensing Rept for Brunswick Steam Electric Plant Unit 2 Reload 12 Cycle 13 ML20249B9691998-06-11011 June 1998 Rev 1 to VC44.F02, Brunswick Steam Electric Plant,Units 1 & 2,ECCS Suction Strainers Replacement Project,Nrc Bulletin 96-003 Final Rept BSEP-98-0129, Monthly Operating Repts for May 1998 for Bsep,Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Bsep,Units 1 & 2 ML20151S9041998-05-31031 May 1998 Revised Pages to Monthly Operating Rept for May 1998 for Brunswick Steam Electric Plant,Unit 1 BSEP-98-0104, Monthly Operating Repts for Apr 1998 for Brunswick Steam Electric Plant,Units 1 & 21998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Brunswick Steam Electric Plant,Units 1 & 2 ML20151S8991998-04-30030 April 1998 Revised Pages to Monthly Operating Rept for Apr 1998 for Brunswick Steam Electric Plant,Unit 1 ML20247N7721998-04-30030 April 1998 Rev 0 to J1103244SRLR, Supplemental Reload Licensing Rept for BSEP Unit 1,Reload 11,Cycle 12 ML20247N7501998-04-30030 April 1998 Rev 0 to BSEP Unit 1,Cycle 12 Colr ML20217K8461998-04-24024 April 1998 Safety Evaluation Approving Proposed Use of Code Case N-535 at Brunswick Unit 1 During Second 10-yr Interval,Pursuant to 10CFR50.55a(a)(3)(i).Authorizes Use of Code Case N-535 Until Code Case Included in Future Rev of RG 1.147 ML20217K3941998-04-24024 April 1998 SER Approving Relief Request for Pump Vibration Monitoring, Brunswick Steam Electric Plant,Units 1 & 2 ML20217E6841998-04-23023 April 1998 Safety Evaluation Accepting Code Case N-547, Alternative Exam Requirements for Pressure Retaining Bolting of CRD Housings ML20217E7471998-04-21021 April 1998 Safety Evaluation Accepting Alternative to Insp of Reactor Pressure Vessel Circumferential Welds ML20217B5241998-04-20020 April 1998 SE Accepting Licensee Request for Approval to Use Alternative Exam Requirement for Brunswick,Unit 1,reactor Vessel Stud & Bushing During Second 10-yr ISI Interval Per 10CFR50.55a(a)(3)(ii) BSEP-98-0080, Monthly Operating Repts for Mar 1998 for Bsep,Units 1 & 21998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Bsep,Units 1 & 2 ML20216B1041998-03-0404 March 1998 SER Approving Alternative to Insp of Reactor Pressure Vessel Circumferential Welds for Brunswick Steam Electric Plant, Unit 1 1999-09-30
[Table view] |
Text
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D6cket Nos. 50-324 October [,1969 and 50-325 1
]
1 Second Supplemental Report to ACR$
BRUNSWICK STEAM ELECTRIC FIANT UNITS 1 & 2 U.S. Atomic Energy Commission Division of Reactor Licenstag p g llN O l' C% V "Lt4IJM= hd* E" ,
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i substantial' leakage in the core standby cooling system (CgCg) suction lines 1 of the Brunswick plant following the design basis loss-of-coolant seeident ;
could lead to the loss of torus water required to provide effective core '
cooling. In our original report to the ACRg, our positium was that the design should be modified by providing isolation valves both inside and out- l side the containment or by other means such as enclosing the section piping i out to and including the first isolation valve. The ACRg ta its letter of May'15, 1969, stated that a short run of pipe, estremely conservative design, remote operability of the first isolation valve, inservice surveillance and leak detection to be a suitable exception to the general rule. We have con- J cluded that the applicant's design approach, as described in Amendment No. 8, 1 does not meet the conditions noted in the ACRg letter. We, therefore, conclude that a suitable design change such as a snard-pipe or jacket sho61d be installed on the core standby cooling system section lines of the Brunswick plant from the torus liner out to and including the first isolation j valve. i I
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1.0 INTRODUCTICII AND SIDMART This Second Supplemental Report to the ACRg discusses the background and bases for our conclusion a design modification such as a guard-pipe or jacket should be installed on the core standby cooling system (CSCs) suction lines for the Brunswick plant from the torus liner out to and including the first isolation valve. This report supplements our previous reportf of September 3,1969 on this subject. This requirement would be consistent with our general approach that, insofar as practicable, multiple protection should be provided for both active and passive failures in systems or com-ponents that must function to protect the health and safety of the public. ;
Cross leakage in the C8CS suction line and/or in its isolation valve following i
a loss-of-coolant accident could potentially lead to a loss of recirculation l d
flow and result in offsite doses that exceed 10 CFR:.100 limits. Properly l l
installed guard-pipes or jackets around the CSCS suction lims from the torus I l
liner out to and including the first isolation valve would provide protection against flailure of the suction lines themselves.
In its May 15, 1969 letter on the Brunswick Plant, the ACRS stated:
" Engineered safety systems that are required to recirculate water after a loss-of-coolant accident should be designed so that a gross system leak will not result in critical loss of recircular*on I or in loss of isolation capability. The Committee believes that exception to this general rule may be made in respect to a very short run of pipe from the torus to the first valve, if extremely con-servative design of the pipe (and its connection to the torus) is used and suitably remote operability of the valve is provided.
The design of these systems also should provide adequate leak detection and surveillance capability."
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We agree with the ACRg that exceptions may be made in the design of the ,
C8CS suction lines provided adequate compensating design features are provided. Se ACRs stated in its letter what design features would, in its view, make the applicant's design concept acceptable. %e applicant's design approach does not provide these features. As a result, we conclude that the use of guard-pipes or other appropriate measures should be required to serve as a second barrier against failure of these lines.
2.0 BASES FOR FROTECTICII AGAINST FASSIVE COHp0 EFT FAILURES In our view, protection against single failures in either active or passive components in the emergency core cooling systems (ECCS) should be required for the long term recirculation cooling phase. h e ECCS must be designed to perform its function throughout the post-seeident {LOCA) recovery period, which may be af long as 6 as 12 months. Significant uncer- -;
tainties exist in the knowledge of potential accident and post-accident conditions and capabilities of the ECCS. %e ECCS should, therefore, be designed to accommodate without serious consequemees a single failure any-where in the system.
Se consequences of e gross failure in any ame of the ECC8 subsystems could be serious if the failure were not promptly isolated. %ese consequences ;
include possible loss of torus water for those pomys in the remaining ECCS subeystems that ary othervi6 afunctional, and possible offsite doses that i l
exceed 10CFR100kimits. %ese two topics are discussed further in the
,gubsequent parageAphs.
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3 la the Brunswick design, as depicted in the attached sketches, the ICCS pumps are located in. five separate compartments, each with its own susp.
3 The two Israest compartments, each with a ecstained volume of about 81,000 ft ,
house the two RER pumps. The sump pump has a 50 spa capacity, but its drive l motor and actuating circuitry is fully dependent on the availability of 'offsite 4 power. A major leak in one of the tsCS section lines (or valves) that could j mot be isolated, would cause torus water' to flow int.o the compartment of the affected subsystem. Since the sump pump capacity is about 50 spa, with ex-ceseive leskage the affeeted eespertmest eould be filled and would paseibly overflow via the compartment stairwell, depending on the torus internal pressure )
The initial (or pre-LOCA) water inventory of the torus is 87,600 f t3.
A torus internal pressure greater than 26 pais could cause the compartment to overflow via the stairwell. Continued leakage could cause the overflow water to spill into the second RER compartment via its stairvell. If this were to occur, it could lead to flooding of the redundant RER sahepstenst
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In addition to the direct effect of flooding, the HPSE requirements of the RER pumps and the core spray pumps which are shown to be 32 and 33 feet,-
respectively, in Table VI-2-4 and VI-2-5 of the PSAR might not be maintained in the event of excessive leakage. The nominal static head from the torus water level is shown to be about 18 feet in Figure V-2-7. To maintain adequate HPSE to avoid cav&tation inPEER and core spray pumps, the torus water tegerature 1
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4 aust be less than about 185'F, or the primary containment must be pressurized at about 7 psi above the saturation pressure of the torus water (the precise value would depend on, the actual terms water level). Since torus. water at a temperature less than 185'F eennot be assured throughout a post-seeident ,
i recovery period,' effective ECCS operation depende en maintaining a positive I
pressureoftheprimarycontainmentunf1pooltemperatureisreduced.
We have also considered potential offsite doses assuming gross leakage in one' 6f the CSCS suction lines, following a LOCA. A number of variables, such as the time af ter the 14CA for. the leakage to occur, the fraction' of torus !
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water 1eaking into the affected C8CS pump coupertment, the torus internal 4 lu.hc9<. _'
pressure at the time of the lokage, and the decontamination factor associated l 8
with the performance of standby gas treatment system and plant stack, sffect significantly the potential offeite doses. However, since leakage from the CSCS lines in effect would be a breach of containment, the resulting offsite i doses. could be well above the 10 CFR 100 guideline doses.
3.0 IVALUAfl g__o_t g BRE L33mtgMicLJtE81 5 !
The ACRS letter dated May 15, 1969, on the grenswick plant cited a number of conditions which, if satisfied, would obviate the need for protection against a leak in the auction Tide of the engineered safety systems. These conditions included: 1) a very short run of pipe from the torus to the first .
l valveg 2) extremely conservative design of the pipe (and its connection to the torus); 3) suitably remote operability of the valves and 4) adequate leak-detection and surveillance capability.
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5 In the applicant's proposed design, the pipe length from terus liner to the first valve is 18 feet, of which 12 feet is through a guide pipe in the concrete wall of the torus and, therefore, inaccessible for inservice inspection. As presently designed, the guide pipe does not prevent or contain leakage from the CSCS piping.
The applicant also proposes an upgrading of the existing USAs 531.1.0 Code of Power Piping and Valves (1967) including additional analysis, and quality control of materials and fabrication in addit &on to that required by the code. We agree that. the proposed additional requirements represent a conservative design. l The first valve on the line between the terms liner and the CSCS pump has been made remotely operable by the ese of a remotely controlled motor driven valve.. We find this meets the requirements as stated in the ACRS letter.
The proposed leak detection system includes the use of levnt switches and sump pumps. New leaks will be detected when the sump pump is observed to operate more frequently and/or for longer duration than usual. The sump pumps and level switches rely on availability of offsite electrical power.
To determine the leakage source, the CSCS isolation valves will be alternately ;
41osed. We find the proposed leak detection system inadequate in terms of ability to detect small leaks and in terms of reliability because the system will be unable to identify the location of leaks that occur in the 18-feet run of pf pe upstr9am of the CSCS suction line isolation valve and the system would be ineffective in the absence of offsite power.
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6 We are not able to condlude that the conditions specified in the ACRS .
letter have been satisfied, because we are not sure that 18 feet should be considered a very short run of pipe, because of the difficulties in inservice inspection, and because of the limitations in the capability for leak detectirra.'
Therefore, we conclude that additional protection should be provided to protect against the consequences of leakage in the CSC8 suction lisas, i -- 4.0 GUARD-FIFE DESIGN _ APPROACB One design approach to permit isolation of pipe b-eaks in lines which penetrate the contalement is to use two valves in series; one inside and one _
outside of containment. In some instances, this design approach is, not consideri practical. The approach chosen by most FWR applicants is to place' a pressure-tight jacket around the ECCS suction lines from the centstanent suay out to and including the first isolation valve. Representative plants using this godsd-pipe approach include Pacific Gas and Electric's Diablo Canyon plant, Sacramento manicipal Utilities District's Rancho seco plant, and the Duke Power ,
Company's ocones plant.
An appropriate guard-pipe design e'en allow access for inservice inspection ,
equal to that which would otherwise be available. Moreover, leak detection can be made anch more effective, with greater sensitivity and reliability, than that of the proposed Branswick design, which uses sump pumps and level switches.
Some of the disadvantages of the guard-pipe approach include possible degradation of the line and valve because of additional welding requirements' and possible differential expansion between the line and the guard-pipe.
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3 .7 Proper design attention is required to minimize this kind of degradation and the design of the line and valve should have sufficient margina to
\ accommodate the residual amount of degradation.
We conclude that the additional protection that would be obtained by use of the guard-pipe, as a recond barrier against possible loss of effective emergency core cooling water and as a second barrier to the release of fission product activity directly from the primary system to the reactor building, outweigh significantly the disadvantages associated with l
potential system degradation. l l
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