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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20140E9181997-06-0707 June 1997 Safety Evaluation Supporting Amend 133 to License DPR-43 ML20247K2531989-09-11011 September 1989 Safety Evaluation Supporting Amends 123 & 41 to Licenses DPR-61 & NPF-49,respectively ML20247E3761989-09-0707 September 1989 Safety Evaluation Supporting Amends 122,34,143 & 40 to Licenses DPR-61,DPR-21,DPR-65 & NPF-49,respectively ML20248C0731989-08-0303 August 1989 Sser Accepting 880601,0909 & 890602 Changes to ATWS Mitigation Sys Actuation Circuitry for Plants ML20246L2571989-06-26026 June 1989 Safety Evaluation Supporting Amends 118,33,142 & 36 to Licenses DPR-61,DPR-21,DPR-65 & NPF-49,respectively ML20245J0751989-04-25025 April 1989 Safety Evaluation Supporting Amends 114,30,141 & 33 to Licenses DPR-61,DPR-21,DPR-65 & NPF-49,respectively ML20205Q5761988-10-31031 October 1988 SER Re Application for CP for Alchemie Facility-2 Oliver Springs.Licensee Technically Qualified to Construct & Operate Proposed Facility in Such Way as to Assure Adequate Protection for Common Defense & Security ML20205Q5721988-10-31031 October 1988 SER Re Const Mod & Licensing of Company Facility-1 Cpdf. Licensee Technically Qualified to Modify Existing Facility in Such Way as to Assure Adequate Protection of Common Defense & Security ML20205M5731988-10-26026 October 1988 Safety Evaluation Supporting Amends 108,25,134 & 26 to Licenses DPR-61,DPR-21,DPR-65 & NPF-49,respectively ML20155G4801988-09-28028 September 1988 Safety Evaluation Supporting Amends 107,23,132 & 24 to Licenses DPR-61,DPR-21,DPR-65 & NPF-24,respectively ML20195G5531988-06-24024 June 1988 Safety Evaluation Accepting Proposed Reracking of Util Spent Fuel Storage Pools from Criticality Standpoint.Enrichment of Fuel to 4.5 Weight % U-235 May Be in Conflict W/ 10CFR51 Table S4 & Should Be Investigated by NRC ML20154H2171988-05-18018 May 1988 Safety Evaluation Accepting Util 880414 Submittal Re Reload Startup Physics Test Program ML20147D9631988-02-25025 February 1988 Safety Evaluation Accepting Util 860404 Evaluation of Environ Qualification of Equipment Considering Superheat Effects of high-energy Line Breaks for Plants,Per IE Info Notice 84-90 ML20147E9261988-02-23023 February 1988 Safety Evaluation Supporting Amends 100,14,125 & 15 to Licenses DPR-61,DPR-21,DPR-65 & NPF-45,respectively ML20236M9631987-11-0606 November 1987 Safety Evaluation Accepting Util Proposed ATWS Mitigating Sys Actuation Circuitry for Facilities,Per 10CFR50.62(c)(1) & Pending Final Resolution of Tech Spec Issue ML20236L2001987-10-30030 October 1987 Safety Evaluation Supporting Amends 11 to Licenses NPF-37 & NPF-66,respectively & Amend 1 to License NPF-72 ML20195G5501987-10-28028 October 1987 Safety Evaluation Accepting Proposed Reracking of Util Spent Fuel Storage Pool.Licensee May Be Required to Perform Addl Analyses to Verify Maint of Required Subcriticality If Staff Boraflex Studies Reveal Gap Development in Subj Boraflex ML20235A7331987-09-18018 September 1987 Safety Evaluation Re Installation of Alternate Rod Injection (ARI) Sys & Adequacy of Plant Reactor Coolant Recirculating Pump Trip (RPT) Sys,In Compliance W/Atws Rule 10CFR50.62. ARI & RPT Acceptable ML20195G5461987-08-21021 August 1987 Safety Evaluation Accepting Proposed Util Tech Specs Change to Increase Max Allowed Enrichment of Reload Fuel to 4.2 Weight % U-235 ML20236Y0221987-07-0808 July 1987 Safety Evaluation Clarifying Determination of Acceptability of Test Duration for Performance of Integrated Leak Rate Test at Plant ML20214M3361987-05-22022 May 1987 Safety Evaluation Supporting Util Rept Entitled, Rod Swap Methodology Rept for Startup Physics Testing ML20214M1491987-05-21021 May 1987 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 2.1 (Part 2) Re Established Interface W/Nsss or Vendors of Each Component of Reactor Trip Sys ML20153F3851987-05-0404 May 1987 SER Accepting Util 870301 & 0407 Requests for Rev to Tech Specs,Modifying MSIV Actuation Control Sys,Per IEEE Std 279 & GDC 21 ML20153F3671987-04-30030 April 1987 Safety Evaluation Accepting Util 870311 Proposed Changes to Tech Spec Tables 2.2.1-1,3.6.1.2-1 & 3.6.3-1 Re MSIV-closure Setpoint ML20207R6481987-03-11011 March 1987 Safety Evaluation Supporting Licensee Response to Generic Ltr 83-28,Item 2.1 (Part 1) Equipment Classification.Program Acceptable ML20212J4041987-03-0202 March 1987 Safety Evaluation Re Util 831104 Response to Generic Ltr 83-28,Item 4.5.2, Reactor Trip Sys Reliability,On-Line Testing. Licensee Meets NRC Position ML20212B5671987-02-24024 February 1987 Safety Evaluation Re Util Response to Generic Ltr 83-28,Item 4.5.2, Reactor Trip Reliability On-Line Testing. Response Acceptable ML20211Q2971987-02-18018 February 1987 Safety Evaluation Re Auxiliary Feedwater Sys Reliability (Generic Issue 124) for Prairie Island Units 1 & 2 ML20211G4361987-02-14014 February 1987 Safety Evaluation Supporting Util 831110 Response to Generic Ltr 83-28,Item 4.5.2 Re on-line Testing for Reactor Trip Sys Reliability ML20153F3511987-02-0707 February 1987 Safety Evaluation Accepting Util Latest Design Mods to MSIV Actuation Control Sys,Per IEEE 279 & GDC 21 ML20210T2571987-02-0606 February 1987 SER Re Util 850802 Submittal Describing Design Details of Steam Generator Blowdown & Auxiliary Steam Sys to Detect & Isolate High Energy Line Breaks.Sys Design Acceptable, However,Two Deviations from IEEE-STD-297 Criteria Apparent ML20210R2061987-02-0606 February 1987 Safety Evaluation Supporting Util 850517,0802,0823,1211 & 860429 Submittals Re Environ Effects of High Energy Line Breaks in Auxiliary Steam or Steam Generator Blowdown Sys. Design of Blowdown Sys Acceptable ML20209B1511987-01-28028 January 1987 SER Supporting Util Responses to Generic Ltr 83-28,Item 4.5.2 Re Reactor Trip Sys on-line Testing ML20209C3571987-01-23023 January 1987 SER Supporting Facility Design,Per Generic Ltr 83-28,Item 4.5.2, Reactor Trip Sys Reliability,On-Line Testing ML20209C2151987-01-21021 January 1987 Safety Evaluation Re Auxiliary Feedwater Sys Reliability (Generic Issue 124) at Prairie Island Units 1 & 2.Util Actively Pursuing Improvements in Sys Reliability & Reducing Sys Challenges ML20215E0061986-12-12012 December 1986 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 2.1 (Part 1) Re Equipment Classification of Reactor Trip Sys Components ML20214J9911986-11-24024 November 1986 SER Re Conformance to Generic Ltr 83-28,Item 2.1 (Part 1) on Equipment Classification (Reactor Trip Sys Components) at Selected GE BWR Plants.Part 1 Requirements Met ML20214T8061986-11-19019 November 1986 Safety Evaluation Supporting Util Response to Generic Ltr 83-28,Item 2.1 Re Equipment Classification of Reactor Trip Sys Components ML20215D7341986-10-0101 October 1986 Safety Evaluation Re Util 860623 Request That One Startup Test Be Modified & Five Startup Tests Be Eliminated.Mod to Rod Drop Measurement Test & Elimination of Certain Other Startup Tests Acceptable ML20214P0181986-09-0909 September 1986 Safety Evaluation Supporting Amends 83,111,112 & 1 to Licenses DPR-61,DPR-21,DPR-65 & NPF-49,respectively ML20214N7201986-09-0909 September 1986 Safety Evaluation Conditionally Supporting Rod Swap Technique & Util Nuclear Analysis Methods for Control Rod Worth Measurements ML20209D3871986-08-28028 August 1986 SER Supporting Util Responses to Generic Ltr 83-28,Item 2.1 (Part 1) Re safety-related Reactor Trip Sys Components ML20153F6291986-08-27027 August 1986 Safety Evaluation Concurring W/Applicant 860703 Exemption Request Re Testing of 16 Relief Valves,Per 10CFR50,App J ML20205H1511986-08-15015 August 1986 Safety Evaluation of Generic Ltr 83-28,Item 2.1 (Part 1) Re Equipment Classification (Reactor Trip Sys Components) ML20236X6941986-08-0505 August 1986 Supplemental Safety Evaluation Supporting Seismic Qualification of Auxiliary Feedwater Sys (Multi-Plant Action C-14).SALP Input & 861029 Fr Re Proposed Rules Section Encl ML20212H7541986-07-24024 July 1986 Safety Evaluation Supporting Licensee 841104,850208 & 0726 Responses to Generic Ltr 83-28,Item 2.1,Part 1, Equipment Classification ML20215A7521986-07-0707 July 1986 SER Supporting Util 860523 Proposed Changes to Tech Specs Re Reactor Trip Sys Instrumentation & Surveillance in Response to Generic Ltr 85-09.SALP Input Also Encl ML20235U6021986-06-17017 June 1986 Sser Supporting Util 860528 Request for Relief from Impractical Preservice Insp Requirements of ASME Code Section XI ML20211F0241986-06-12012 June 1986 Safety Evaluation Supporting Util Listed Responses to Generic Ltr 83-28,Item 2.1 (Part 1) Re Identification & Classification of Reactor Trip Sys Components IA-87-213, SER Finding Facility Design Acceptable W/Respect to Conformance to Reg Guide 1.97,Rev 3,except for Accumulator Tank Level & Pressure Instrumentation,Containment Sump Water Temp Instrumentation & Quench Tank Temp Instrumentation1986-06-11011 June 1986 SER Finding Facility Design Acceptable W/Respect to Conformance to Reg Guide 1.97,Rev 3,except for Accumulator Tank Level & Pressure Instrumentation,Containment Sump Water Temp Instrumentation & Quench Tank Temp Instrumentation 1997-06-07
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20140E9181997-06-0707 June 1997 Safety Evaluation Supporting Amend 133 to License DPR-43 ML20115J3981996-07-22022 July 1996 Interim Part 21 Rept Re 3/4 Schedule 80 Pipe Furnished to Consolidated Power Supply.Investigation Revealed Only One Nuclear Customer Involved in Sale of Matl ML20126J5961992-12-31031 December 1992 Part 21 Rept Re Potential Loss of RHR Cooling During Nozzle Dam Removal.Nozzle Dams May Create Trapped Air Column Behind Cold Leg Nozzle Dam.Mod to Nozzle Dams Currently Underway. Ltrs to Affected Utils Encl ML20113H4871992-07-30030 July 1992 Part 21 Rept Re Surge Line Flooding Which May Occur When Large Pressurizer Vent Used to Support Reduced Rc Inventory Operations.Notification Received from Westinghouse on 920312.Item Not Reportable Per 10CFR21 ML20086U0311991-11-30030 November 1991 SG Tube Plugging Criteria for Outside Diameter Stress Corrosion Cracking (ODSCC) at Tube Support Plates ML20077K0831991-06-30030 June 1991 Nonproprietary Technical Justification for Eliminating Pressurizer Surge Line Rupture as Structural Design Basis for Kewaunee Nuclear Plant ML1118102861991-05-31031 May 1991 Handbook on Flaw Evaluation Kewaunee,Unit 1 Steam Generators Upper Shell to Cone Weld. ML20091E3441991-03-31031 March 1991 Nonproprietary Structural Evaluation of Kewaunee Pressurizer Surge Line,Considering Effects of Thermal Stratification ML20066B2101990-12-31031 December 1990 Generator Mid-Cycle Rept,Dec 1990 ML20062A4641990-09-30030 September 1990 Nonproprietary, Reevaluation of U-Bend Tube Fatique for Kewaunee Plant Steam Generators ML20042F3261990-05-0404 May 1990 LER 90-005-00:on 900404,intergranular Attack & IGSCC Results in Both Steam Generators a & B Being Categorized as C-3.All Defective Tubes Plugged & Sludge Lancing & Crevice Flushing conducted.W/900504 Ltr ML1117803551990-04-30030 April 1990 Generator Tubesheet Crevice Indications Return to Power Rept. ML18094B3211990-02-28028 February 1990 Annual Operating Repts for 1989 for Salem & Hope Creek Generating Stations ML20012A9011990-02-27027 February 1990 Suppls 900213 10CFR21 Rept Re Chilled Water Sys Operation. Evaluation of Crystal River Determined That Postulated High Energy Line Break in Intermediate Bldg May Be Subj to Steam Loads Higher than Normal Loads,Causing Rising Water Temp ML20011F5971990-02-22022 February 1990 Part 21 Rept Re Solder Connections in Abb 27/59 Relays Deteriorated Due to Thermal Stress,Causing Bonding of Printed Wiring Pattern to Glass Epoxy Circuit Board.Interim Circuit Board W/Larger Pads & Higher Wattage Will Be Used ML20011F1941990-02-22022 February 1990 Part 21 Rept Re Abb 27/59 Relay Catalog Series 211L.Solder Connections to Printed Wiring Runs on Bottom of Circuit Board Deteriorated Due to Thermal Stress.No Actual Failure Occurred & Relays to Be Changed at Next Outage ML18153C1011990-02-0202 February 1990 Part 21 Rept Re Two of Three Pc Cards in GE Type SLV11A1 Over/Undervoltage Relays Failing to Produce Output.Short Between Leads Would Result in Damage to Component 1C5. Sketch of Threshold Detection Board Encl ML17223A7451990-01-26026 January 1990 Part 21 Rept Re Backup Rings Furnished in Spare Parts Seal Kits & in 25 Gpm 4 Way Valves as Part of Actuators Made of Incorrect Matl.Rings Should Be Viton & Have Been Identified as Buna N ML20006A8231990-01-10010 January 1990 Errata to Rev 3 to BAW-1543, Master Integrated Reactor Vessel Surveillance Program Consisting of Revised Tables 3-20 & E-1 ML20005G6831990-01-0505 January 1990 Part 21 Rept Re Installation Instructions for Grommet Use Range for Patel Conduit Seal P/N 841206.Conduit Seals in Environ Qualification Applications Inspected for Proper Wire Use Range & Grommets Replaced ML17347B4621989-12-31031 December 1989 App a to USI A-46 & Generic Ltr 87-02. ML19354D8351989-11-0808 November 1989 Presentation to NRC Re Kewaunee (Wps) SECY-83-472 LOCA Analysis Effort. ML18094B1471989-10-25025 October 1989 Emergency Plan Annual Exercise 1989 for Artificial Island on 891025. W/One Oversize Drawing ML19325E0861989-10-16016 October 1989 Followup Part 21 Rept Re Class 1E Battery Chargers W/ Transformers Running at Temps Exceeding Those Used in Qualification Rept When Operating at or Near Full Load Rating of Equipment.Listed Corrective Actions Underway ML19351A2941989-10-0909 October 1989 Part 21 Rept Re Potential of Ambient Compensated Molded Case Circuit Breakers to Deviate from Published Info. Instantaneous Trip Check Will Be Instituted on All Class 1E Thermal/Magnetic Ambient Breakers Prior to Shipment ML20248G8291989-10-0202 October 1989 Rev 19 to YOQAP-I-A, Operational QA Program ML17347B3821989-09-30030 September 1989 Monthly Operating Repts for Sept 1989 for Turkey Point Units 3 & 4 & St Lucie Units 1 & 2.W/891016 Ltr ML19351A4191989-09-30030 September 1989 Mark-BW Reload LOCA Analysis for Catawba & McGuire Units. ML19327C0681989-09-30030 September 1989 Nuclear Safety & Compliance Semiannual Rept Number 11,Apr- Sept 1989. W/891027 Ltr ML20248F0001989-09-29029 September 1989 Debris in Containment Recirculation Sumps, Technical Review Rept ML19325C9521989-09-29029 September 1989 Part 21 Rept Re Potential Common Failure of SMB-000 & SMB-00 Cam Type Torque Switches Supplied Prior to 1981 & 1976. Vendor Recommends That Switch W/Fiber Spacer Be Replaced ML20248E0121989-09-13013 September 1989 Supplemental Part 21 Rept Re Potential Problem W/Six Specific Engine Control Devices in Air Start,Lube Oil, Jacket Water & Crankcase Sys.Initially Reported on 890429. California Controls Co Will Redesign Valve Seating ML20248D1571989-09-13013 September 1989 Rev 56 to QA Program ML20247K2531989-09-11011 September 1989 Safety Evaluation Supporting Amends 123 & 41 to Licenses DPR-61 & NPF-49,respectively ML20247E3761989-09-0707 September 1989 Safety Evaluation Supporting Amends 122,34,143 & 40 to Licenses DPR-61,DPR-21,DPR-65 & NPF-49,respectively ML17347B3341989-08-31031 August 1989 Monthly Operating Repts for Aug 1989 for Turkey Point Units 3 & 4 & St Lucie Units 1 & 2.W/890913 Ltr ML20246D6871989-08-14014 August 1989 Rev 1 to Criticality Analysis of Byron & Braidwood Station High Density Fuel Racks ML20248C0731989-08-0303 August 1989 Sser Accepting 880601,0909 & 890602 Changes to ATWS Mitigation Sys Actuation Circuitry for Plants ML19327B4011989-07-31031 July 1989 Safety Evaluation for Byron/Braidwood Stations Units 1 & 2 Transition to Westinghouse 17 X 17 Vantage 5 Fuel. ML18008A0311989-07-31031 July 1989 NTH-TR-01 Decrease in Heat Removal by Secondary Sys. ML17347B2731989-07-31031 July 1989 Monthly Operating Repts for Jul 1989 for Turkey Point Units 3 & 4 & St Lucie Units 1 & 2 ML20246P7111989-07-17017 July 1989 Part 21 Rept Re Quench Cracks in Bar of A-SA-193 Grade B7 Component.Quench Cracks Found in One Bar of Matl.Listed Purchasers Informed of Potential Defect.Next Rept Will Be Submitted When Addl Info Becomes Available ML20247D3011989-07-12012 July 1989 Part 21 Rept 10CFR21-0047 Re Control Wiring Insulation of Inner Jacket Used on General Motors Diesel Generator Sets Identified as 999 or MP Series.Encl List of Owners of Units Notified ML20246D6711989-06-30030 June 1989 Criticality Analysis of Byron/Braidwood Fresh Fuel Racks ML17347B1851989-06-30030 June 1989 Monthly Operating Repts for June 1989 for St Lucie Units 1 & 2 & Turkey Point Units 3 & 4.W/890717 Ltr ML17347B2741989-06-30030 June 1989 Corrected Monthly Operating Repts for June 1989 for Turkey Point Units 3 & 4 ML20247H0711989-06-30030 June 1989 Description & Verification Summary of Computer Program, Gappipe ML20246L2571989-06-26026 June 1989 Safety Evaluation Supporting Amends 118,33,142 & 36 to Licenses DPR-61,DPR-21,DPR-65 & NPF-49,respectively ML20247H0791989-06-22022 June 1989 App to Description & Verification Summary of Computer Program,Gappipe ML18151A5411989-06-21021 June 1989 Updated Operational QA Program Topical Rept. 1997-06-07
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CONTAINMENT SYSTEMS BRANCH g EVALUATION Of TECHNICAL SPE CI F I C AT' ION F OR CONTAINMENT
, COOLING S YJ T E M , ,
KEWAUNEE NUCLEAR POWER PLANT -
DOCKET NO. 50-3D5 .,
A recent review of the Kewaunee Nuclear Power Plan,t Technical Specifications has uncovered deficiencies in the Technical Specification regarding containment cooling system operability. i 5ection 6.3.1 of the Kewaunee Updated final Safety Analysis Report (UI S AR) defines the minimum allowable containment cooling capability for accident conditions as follows: '
Any of the following combinations of equipment will provide sufficient heat removal capability to maintain' the post-accident containment pressure below'the design value.
- a. All four containment fan-coil units,
- b. Two containment spray pumps,
- c. Two of the four fan-coit units and one containment spray pump.
Portions of other systems which share functions and become part of t his containment cooling system are designed to meet the criteria of this section. No active failure will de-1 grade the heat removal capability of the containment ~ cooling function to less~than 100% of that required for the post !
l loss-of-coolant accident. i i
8707090271 870701 PDR FOIA Wil.LI AMB7-122 PDR i
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4 i
14 1 I
Section 6.3.3 of the UFSAR e.s t ab lJ s h e s- th e, ba si s for the minimum
. operability of the containment cooling systen as folloss:
~
The minimum equipment required to provide sufficient heat removal capability to maintain the post-accident containment pressure below the design valuer assuming that the core' residual heat is released to the contain- -
rent es steam and water is: -
.two containment fan-coil units, and i
l one containment spray pump.
However, T e c hni c a l Spe'c i f i c a t ion 3.3 b.2 re a ds as fo'llows:
- 2. During power operation or recovery from inadvertent trip, any one of the following conditions of inopera-bility may exist during the time intervals specified.
The reactor shall be placed in the hot shutdown condition if operability is not restored within the tine specified, and it shall be placed in the cold shutdown condition i f operability is not restored within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
A. ONE fan-coit unit may be out of service for.a .
4 period not to exceed 7 days.
B. ONE containment spray pump and/or its flow-path may l be out of service for a p6rio6 not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided t he remaining containment spray pump and its flow path are operable, prior t o initiating maint enance and provided three fan-coit units are demonstrated to be operable.
a
~
. ~ . . ,-
This Technical Specification is deficient in t hat it allows taking
. . ~ ~~. -
a fan-coil out of service for 7 days, and not demonst' rating opera-
.. I bili,ty of the remaining fan-coils or the containment spray pumps.
It also allows taking a containment spray out of service while one 1sn-coil'is out of service with no demonstration of the operability of the remaining spray pump. With one fan-coil and one spray pump out of service and the single failure of the remaining sprey pump, the plant is left with only 3 fan-coils to preve6t overpressuriza- ;
1 tion of the containment. This is not in accordance with the design basis as stated in the UFSAR.
- T e c hni c a l Spe ci fi c at i o,n 3.3.b.2 s hould be modi f i ed t o correct these deficiencies. The following is provided as a possible replacement specification for 3.3.b.2; marginal bars are provided l 1
'to indicate the suggested changes:
- 2. During power operation or recovery from inadvertent l trip, any one of the following conditions of inopera-bility may emist during the time i nt e r va' I s s p e c'i f i e d.
i The reactor shall be placed in the hot shutdown condi-tion if operability is not restored within the time specified, and it shall be placed in the cold shutdown condition if operability is not restored within an
, additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. -
A. ONE fan-coit unit may be out of service for a period j not to exceed 7 days. The remaining fan-coit units l
shall be t est ed to demonst rat e operability before ini- l tiating maintenance on the inoperable fan-coit unit j
v 7 . - .
-' /
+ _ . ,
and shall be tested once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,thereafter .,
until all fan-coit units are in an'oper$ble status or the reactor is shut down, i
B. ONE containment spray pump and/or its flow path may be out of service for a period not to exceed 4E. hours.
The rer.aining containment spray pOnp and its flow path shall be tested to dec.onstrate operability before initiating main'tenance on the inoperable pump or flow path and shall be tested once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> there-after, until both pumps and their flow paths are .in an operable status or the reactor is shut down.
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