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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20210K4721999-08-0303 August 1999 SER Accepting Licensee 180-day Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20206A2111999-04-21021 April 1999 Safety Evaluation Authorizing Pump Relief Request PRR-001 & Valve Relief Request VRR-001 & Denying Valve Relief Request VRR-002 ML20205D5481999-03-26026 March 1999 SER Accepting Util Proposed Alternative to Exam Weld AA with Weld Volume Coverage of 62 Percent for First 10-year Insp Interval Pursuant to 10CFR50.55a(a)(3)(ii) & 10CFR50.55a(g)(6)(ii)(A)(5) 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 ML20155H3491998-11-0303 November 1998 Safety Evaluation Granting Requests for Relief RR2-0001, RR2-0002 & RR2-0003 ML20151S5421998-09-0303 September 1998 Safety Evaluation Opposing Licensee Thermal Model as Currently Implemented.Evaluation Recommended to Be Used in Any follow-up Site Insp ML20249A6431998-06-12012 June 1998 SER Accepting 980427 Request for Change to River Bend QA Manual Program Description,Per 10CFR50.54(a)(3) 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) ML20202F0191997-11-28028 November 1997 Safety Evaluation Approving Transfer of License NPF-47 for River Bend Station,Unit 1 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 ML20138C8501997-04-28028 April 1997 Safety Evaluation Authorizing Proposed Alternative to Vibration Measurement Requirements of OM-6,paragraph 4.6.4 (B),For Standby Svc Water Pumps & Diesel Fuel Oil Transfer Pumps,Per 10CFR50.55a(a)(3)(ii) ML20147H4371997-03-31031 March 1997 Safety Evaluation Accepting Proposed P-T Limits for Heatup, Cooldown,Inservice Leak & Hydrostatic Testing & Core Criticality 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 ML20107F5611996-04-17017 April 1996 Safety Evaluation Providing Guidance on Submitting plant- Specific Info W/Respect to IST Program Alternatives Request ML20149E8831994-08-0202 August 1994 Safety Evaluation Accepting Interim Relief Request IRR-03 Re Drywell Isolation Check Valves in Equipment Drain Lines & Reactor Equipment Closed Cooling Water Sys ML20058H0221993-11-30030 November 1993 Safety Evaluation Accepting Licensee Proposed Alternative to ASME Code Requirements in PRR 15,per 10CFR50.55(a)(3)(i). Denies Pump Relief Request 14 ML20059B2771993-10-22022 October 1993 Safety Evaluation Accepting IST Program Relief Requests ML20056G2001993-08-19019 August 1993 Safety Evaluation Denying 930419 Request for Relief from Requirements for Hydrostatic Testing of ASME Class 3 Pressure Retaining Components ML20126A2971992-12-16016 December 1992 Safety Evaluation Accepting Design for Drywell Hydrogen Concentration & Suppression Pool Water Level,W/Respect to Reg Guide 1.97 ML20128D2251992-12-0202 December 1992 Safety Evaluation Accepting Methodology Described in Topical Rept EA-CA-91-0001-M, Steady State Core Physics Methods for BWR Designs & Analysis ML20055F1891990-06-27027 June 1990 Safety Evaluation Supporting Util Tech Spec Change Request Re Scram Accumulator Check Valves ML20245A1051989-06-15015 June 1989 Safety Evaluation Accepting Util 890515 & 19 Submittals Re Results of Augmented IGSCC Insp & Flaw Evaluation of Feedwater Inlet Nozzle N4-A to Safe End Weld ML20247J7411989-05-25025 May 1989 Safety Evaluation Accepting Existing Intervals for on-line Functional Testing for Achieving High Reactor Trip Sys Reliability at All Operating Reactors,Per Generic Ltr 83-28, Item 4.5.3 ML20235N6031989-02-21021 February 1989 Safety Evaluation Approving Alternate Method for Complying w/10CFR50.55a for Main Steam Positive Leakage Control Sys Valve ML20195H1511988-11-23023 November 1988 Safety Evaluation Accepting Responses Re Item 4.5.2 of Generic Ltr 83-28 Re Reactor Trip Sys Reliability on-line Testing ML20196A6611988-11-18018 November 1988 SER Accepting Util Response to Generic Ltr 83-28,Item 2.1, Part 2 Re Vendor Interface Programs for Reactor Trip Sys Components ML20195G8861988-11-18018 November 1988 SER Accepting Util Responses to Generic Ltr 83-28,Item 2.1, Part 1 Re Reactor Trip Sys Components ML20154M6561988-09-22022 September 1988 Safety Evaluation Supporting Licensee post-trip Review Program & Procedures Per Generic Ltr 83-28,Item 1.1 Re Generic Implications for Salem ATWS Events ML20151Q4701988-08-0101 August 1988 Supplemental Safety Evaluation Re First 10-yr Inservice Insp Program.Nrc Concurs W/Determination Made by Licensee That Section XI of ASME Code Exam Requirements for Category B-A Welds in Reactor Pressure Vessel Are Impractical to Perform ML20150F9501988-07-12012 July 1988 Safety Evaluation Supporting Util 840803 & 870922 Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1 & 3.2.2 Re post-maint Testing of All safety-related Components ML20148D4391988-01-14014 January 1988 Safety Evaluation Supporting Util 870731 & 1030 Submittals on Facility Compliance Re Alternate Rod Injection & Recirculation Pump Trip Sys,Per ATWS Rule 10CFR50.62 ML20236D1851987-10-20020 October 1987 Safety Evaluation Accepting Util 870602 Rev 2 to First 10-yr Interval Inservice Insp Program ML20235K6871987-09-28028 September 1987 Safety Evaluation Accepting Util Monitoring Sys for UHS as Proposed in NUREG-0989, Safety Evaluation Supporting Proposed Turbine Maint Program. Proposal Satisfies Licensee Commitment Prescribed in Sser 5. Commitment Acceptable Resolution of Turbine Maint Issue1987-08-26026 August 1987 Safety Evaluation Supporting Proposed Turbine Maint Program. Proposal Satisfies Licensee Commitment Prescribed in Sser 5. Commitment Acceptable Resolution of Turbine Maint Issue ML20209B4291987-04-22022 April 1987 Safety Evaluation Supporting Util Proposed Rev to Tech Specs Re Drywell Leakage Test Interval ML20234E0491987-03-26026 March 1987 SER Accepting Response to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Re post-maint Testing of Reactor Trip Sys & All Other safety-related Components ML20205M3691987-03-26026 March 1987 Sser Re Mark III Containment Related Issues ML20212L2351987-01-21021 January 1987 Sser Re Util 860512 & 1119 Rev 4 to Preservice Insp Program & Relief Request R 0012 (a to E) from Certain ASME Code Requirements.Relief Granted.Preservice Insp Program in Compliance w/10CFR50.55a(g)(3) ML20203N7061986-10-0909 October 1986 Safety Evaluation Supporting Amend 2 to License NPF-47 ML20214P2061986-09-19019 September 1986 SER Supporting 860917 & 19 Requests for Amend to License NPF-40,changing Tech Spec Table 3.6.4-1 to Allow Inoperability of Inboard Isolation Valve MOVFO76 Through 861004 ML20207H4151986-07-16016 July 1986 Safety Evaluation Supporting Technical Solution to Address Operability & Reliability of Tdi Emergency Diesel Generators ML20133F8241985-08-0505 August 1985 Marked-up Draft Sser 2 Re Application for Ol,Reporting Status of Certain Items Not Resolved at Time of SER Publication 1999-08-03
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARRBG-45144, Monthly Operating Rept for Sept 1999 for River Bend Station. with1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for River Bend Station. with ML20216G1201999-09-0909 September 1999 Rev 3 to Rbs,Cycle 9 Colr RBG-45110, Monthly Operating Rept for Aug 1999 for River Bend Station, Unit 1.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for River Bend Station, Unit 1.With RBG-45087, Special Rept:On 990510,LPCI a & LPCS Injected Into Rv for Less than Two Minutes.Caused by Electrical Transient in One of ECCS Power Supplies.Operators Verified Reactor Cavity Level & Closed Injection Valves to Stop Injection1999-08-0606 August 1999 Special Rept:On 990510,LPCI a & LPCS Injected Into Rv for Less than Two Minutes.Caused by Electrical Transient in One of ECCS Power Supplies.Operators Verified Reactor Cavity Level & Closed Injection Valves to Stop Injection ML20210K4721999-08-0303 August 1999 SER Accepting Licensee 180-day Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves RBG-45091, Monthly Operating Rept for July 1999 for River Bend Station, Unit 1.With1999-07-31031 July 1999 Monthly Operating Rept for July 1999 for River Bend Station, Unit 1.With ML20210C6391999-07-0202 July 1999 Rev 2 to River Bend Station,Cycle 9 Colr RBG-45055, Monthly Operating Rept for June 1999 for River Bend Station, Unit 1.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for River Bend Station, Unit 1.With ML20196J8031999-06-24024 June 1999 Rev 1 to Rbs,Cycle 9 Colr RBG-45028, Monthly Operating Rept for May 1999 for River Bend Station, Unit 1.With1999-05-31031 May 1999 Monthly Operating Rept for May 1999 for River Bend Station, Unit 1.With RBG-45016, Monthly Operating Rept for Apr 1999 for River Bend Station, Unit 1.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for River Bend Station, Unit 1.With ML20206A2111999-04-21021 April 1999 Safety Evaluation Authorizing Pump Relief Request PRR-001 & Valve Relief Request VRR-001 & Denying Valve Relief Request VRR-002 RBG-44969, Monthly Operating Rept for Mar 1999 for River Bend Station, Unit 1.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for River Bend Station, Unit 1.With ML20205S0601999-03-31031 March 1999 Rept on Status of Public Petitions Under 10CFR2.206 with Status Change from Previous Update,990331 ML20205D5481999-03-26026 March 1999 SER Accepting Util Proposed Alternative to Exam Weld AA with Weld Volume Coverage of 62 Percent for First 10-year Insp Interval Pursuant to 10CFR50.55a(a)(3)(ii) & 10CFR50.55a(g)(6)(ii)(A)(5) RBG-44930, Monthly Operating Rept for Feb 1999 for River Bend Station, Unit 1.With1999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for River Bend Station, Unit 1.With RBG-44826, Monthly Operating Rept for Dec 1998 for River Bend Station Unit 1.With1998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for River Bend Station Unit 1.With ML20198K2701998-12-22022 December 1998 Rev 1 to RBS Cycle 8 Colr RBG-44773, Monthly Operating Rept for Nov 1998 for River Bend Station, Unit 1.With1998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for River Bend Station, Unit 1.With RBG-44727, LER 98-S01-00:on 981021,identified That Contract Employee with Temporary Access Authorization Failed to Disclose Complete Criminal & Employment History.Individual Denied Access to Plants for Five Years.With1998-11-17017 November 1998 LER 98-S01-00:on 981021,identified That Contract Employee with Temporary Access Authorization Failed to Disclose Complete Criminal & Employment History.Individual Denied Access to Plants for Five Years.With 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 ML20155H3491998-11-0303 November 1998 Safety Evaluation Granting Requests for Relief RR2-0001, RR2-0002 & RR2-0003 RBG-44719, Monthly Operating Rept for Oct 1998 for River Bend Station, Unit 1.With1998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for River Bend Station, Unit 1.With ML20155C1351998-10-26026 October 1998 Rev B to Entergy QA Program Manual RBG-44677, Monthly Operating Rept for Sept 1998 for River Bend Station, Unit 1.With1998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for River Bend Station, Unit 1.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 ML20151S5421998-09-0303 September 1998 Safety Evaluation Opposing Licensee Thermal Model as Currently Implemented.Evaluation Recommended to Be Used in Any follow-up Site Insp RBG-44629, Monthly Operating Rept for Aug 1998 for River Bend Station, Unit 1.With1998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for River Bend Station, Unit 1.With 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 RBG-44600, Monthly Operating Rept for July 1998 for River Bend Station1998-07-31031 July 1998 Monthly Operating Rept for July 1998 for River Bend Station RBG-44564, Monthly Operating Rept for June 1998 for River Bend Station, Unit 11998-06-30030 June 1998 Monthly Operating Rept for June 1998 for River Bend Station, Unit 1 ML20249A6431998-06-12012 June 1998 SER Accepting 980427 Request for Change to River Bend QA Manual Program Description,Per 10CFR50.54(a)(3) RBG-44539, Monthly Operating Rept for May 1998 for River Bend Station, Unit 11998-05-31031 May 1998 Monthly Operating Rept for May 1998 for River Bend Station, Unit 1 RBG-44501, Monthly Operating Rept for Apr 1998 for River Bend Station, Unit 11998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for River Bend Station, Unit 1 ML20217M8951998-04-30030 April 1998 QA Program Manual 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) RBG-44458, Monthly Operating Rept for Mar 1998 for River Bend Station, Unit 11998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for River Bend Station, Unit 1 RBG-44423, Monthly Operating Rept for Feb 1998 for River Bend Station Unit 11998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for River Bend Station Unit 1 ML20203L6631998-02-11011 February 1998 Rev 0 to River Bend Station,1998 Emergency Preparedness Exercise RBG-44385, Monthly Operating Rept for Jan 1998 for River Bend Station, Unit 11998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for River Bend Station, Unit 1 RBG-44353, Monthly Operating Rept for Dec 1997 for River Bend Station, Unit 11997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for River Bend Station, Unit 1 RBG-44460, Forwards Annual Radiological Environmental Operating Rept, for Period Jan-Dec 19971997-12-31031 December 1997 Forwards Annual Radiological Environmental Operating Rept, for Period Jan-Dec 1997 ML20203H9891997-12-12012 December 1997 Part 21 Rept Re Potential Manufacturing Defect of Enterprise DSR-4 & DSRV-4 Edgs.Cooper Energy Svcs Supplied 1A-7840 Adjusting Screw to Affected Utils & Sites.River Bend Station Replaced W/Acceptable Assemblies ML20202E9941997-12-0101 December 1997 ISI Plan Second Ten-Yr Interval (Dec 1,1997-Nov 30,2007) ML20199K9741997-11-30030 November 1997 Brief Aerial Photography Analysis of RBS at St Francisville,LA:1996-1997 RBG-44337, Monthly Operating Rept for Nov 1997 for River Bend Station, Unit 11997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for River Bend Station, Unit 1 ML20202F0191997-11-28028 November 1997 Safety Evaluation Approving Transfer of License NPF-47 for River Bend Station,Unit 1 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 RBG-44295, Monthly Operating Rept for Oct 1997 for River Bend Station Unit 11997-10-31031 October 1997 Monthly Operating Rept for Oct 1997 for River Bend Station Unit 1 RBG-44243, LER 97-S03-00:on 970911,contract Employee Was Granted Temporary Access After Failing to List Prior Employment Termination & Access Denial.Caused by Employee Failure to Provide Accurate Info.Updated Info Re Employee1997-10-13013 October 1997 LER 97-S03-00:on 970911,contract Employee Was Granted Temporary Access After Failing to List Prior Employment Termination & Access Denial.Caused by Employee Failure to Provide Accurate Info.Updated Info Re Employee 1999-09-09
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$;f gC SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 2 TO FACILITY OPERATING LICENSE NO. NPF-47 GULF STATES UTILITIES COMPANY RIVER BEND STATION, UNIT 1 DOCKET NO. 50-458
1.0 INTRODUCTION
In River Bend Technical Specification 3.6.4, Table 3.6.4-1, Item a.1 identifies the reactor core isolation cooling (RCIC) steam supply valve 1E51*MOVF076 (M0VF076) as a primary containment automatic isolation valve.
By letters dated September 17, 1986 and September 19, 1986, the Gulf States Utilities Company (GSU) requested a one-time change to the Table by adding a footnote, applicable to isolation valve MOVF076, that would exempt the operability requirements for this valve from September 22, 1986 through October 4, 1986.
While the reactor was at 100% power on September 8, 1986 the M0VF076 valve was undergoing a functional test of the annunciator when the MOV76 breaker tripped due to a ground in the system. A subsequent attempt to stroke the valve indicated the valve to be fully open. As required by the River Bend Technical Specification 3.6.4 Action Statements, the outboard isolation valve 1E51*M0VF064 (M0VF064) was closed. Consequently, the RCIC system was rendered inoperable and by Specification 3.7.3, the plant was placed in a 14-day Action Statement that requires the plant to initiate a shutdown on September 22, 1986.
The GSU proposed technical specification revision will exempt valve MOVF076 from technical specification operability requirements, thereby allowing isolation valve F064 to be opened and returning the RCIC system to operable status. This reouest would be a one-time modification to the Technical Specification effective through October 4, 1986, at which time GSU has scheduled a shutdown and will repair the inoperable isolation valve (M0VF076).
2.0 EVALUATION The RCIC steam turbine supply line is an 8-inch line that branches off the main steamline "A". Inside the River Bend drywell the 8-inch line has an inboard isolation valve 1E51*M0VF063 (M0VF063) and circumventing this valve is a 3/4 inch bypass line consisting of a manual isolation valve and the motor-operated isolation valve M0VF076. The 3/4-inch bypass line is parallel to the 8-inch inboard isolation valve; its functions h id1No2 b e65'009 ' l PDR ADOCK 05000458
- P _ ___ .- - PDR .. - J
w during start-up are to equalize the pressure around MOV63 and to warmup the RCIC steamline. The common out-board isolation valve (8-inch) is MOVF064. During power operation, the normal RCIC turbine steam supply line valve lineup is M0VF076 closed and inboard and outboard isolation valves MOVF063 and MOVF064 open.
The RCIC is a high pressure system primarily designed to maintain suffi-cient water in the reactor pressure vessel to cool the core and to maintain the reactor in the standby condition in the event the vessel becomes isolated and feedwater is not available. During a LOCA, RCIC (which augments other core injection systems) initiates on low vessel water level and delivers rated flow to the vessel through a connection in the vessel head.
1 Even though RCIC is not a part of the ECCS network, the RCIC system is designated as a safety-related system. During limiting conditions of operation (LCO) (i.e., when HPCS is inoperable), power operation is allowed to continue for a period of time provided RCIC is operable.
Valve MOVF076 is only used for RCIC turbine steam supply line warm-up before start-up of the system and the ability to close the valve is not required for the RCIC system to perform its intended design function.
Therefore, in the open position MOVF076 will not affect system operation.
The RCIC steam line isolation valves MOVF063 and M0VF064 are normally kept open and should be kept open in order to keep the steam inlet lines warmed-up and to minimize the potential for water hammer upon system startup. Accordingly, the RCIC steam supply line does not receive the standard containment isolation signals (e.g., high drywell pressure and low reactor water level). The isolation signals that serve the inboard and outboard isolation valve monitor conditions to reveal a break in the RCIC system.
The applicant in a letter dated September 19, 1986, also submitted a radiation dose assessment. In the event of a pipe break downstream of the outboard isolation valve (MOVF064), the leakage will be detected by the leakage detection systems. The leakage detection systems are the area i temperature monitoring system and the RCIC steam line flow monitoring system. Either leakage detection system will provide an isolation signal to MOVF063 and M0VF064. Once a signal is received, a total of a 20-second (a 10-second delay for diesel start plus a 10-second delay for valve closure) delay is assumed for closing of the valve. Full flow is assumed through the 8-inch line for the first 20 seconds. At the end of 20 seconds, the inboard-isolation valve F063 is assumed to close, but the outboard isolation valve fails to close (single failure). With the inoperable valve, F076, in the failed open position and the outboard F064 assumed to be failed open, there will be an unisolated leakage path (3/4-inch line) from the reactor to the auxiliary building. The maximum expected flow through the bypass .(F076) is 30 GPM. Feedwater flow could provide make-up for this fluid loss. If feedwater is not available, the
{
l l
l High Pressure Core Spray (HPCSI system could be used to provide cooling water to the vessel, thereby preventing uncovering of the reactor fuel.
Fuel failure would, therefore, be precluded. The only radioactivity available for release from the break is that which is present in the reactor coolant prior to the break. The bypass flow is assumed to be constant for the first 30 minutes of the transient and then is ramped down to zero in the next 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> as the reactor is brought down to a cold shutdown condition. The total mass released from the RCIC steam line break is estimated to be about 50,000 lbs compared with the-140,000 lbs estimated by the staff for Main Steam Line Break Accident Analysis. In section 15.6.4 of the River Bend SER, we concluded that.the doses for Main Steam Line Break are only a small fraction of 10 CFR Part 100 Guideline values. Since the RCIC piping break with unisolated bypass flow through the F076 valve is bounded by the Main Steam Line Break Analysis, the expected doses for the RCIC break also will be a small fraction of 10 CFR Part 100 guideline values, and hence within our acceptance criteria.
The staff concurs with GSU's assessment that there is no significant reduction .in the margin of safety relating to the temporary period (i.e.,
about 12 days) of having the M0VF076 valve in an inoperable condition.
The alternative approach would require reactor shutdown in order to enter the drywell and subject plant personnel to exposures from the harsh drywell environment. Therefore, such an action would induce unwarranted and inherent challenges to various plant systems and to the affected plant personnel.
3.0 BASIS FOR EMERGENCY CIRCUMSTANCES The Commission published guidance concerning emergency circumstances in the Federal Register on March 6, 1986 (51 FR 7762) which states, in part:
"The staff may receive an amendment request and find an emergency situation, where failure to act in a timely way would result in derating or shutdown of a nuclear power plant...(NRC) may proceed to issue the license amendment, if it determines, among other things, that no significant hazards considerations are involved." In addition, the Commission further states in 51 FR 7762, "...that it expects its licensees to apply for license amendments in a timely fashion..."
On September 8, 1986, while the licensee was performing a functional test on annunciator 1E51*M0VF076, the MOV76 breaker tripped with the valve open.
MOV76 is a 3/4" bypass line valve used for prewarming the RCIC steam line.
With this valve open RCIC was declared inoperable and the licensee entered into a 14-day LC0 within which time the licensee had to restore RCIC to operable status or shutdown.
The licensee determined that the breaker malfunction was due to an unintended electrical ground, apparently due to a wet valve motor. Since the valve is inside the containment, the licensee attempted to close the valve by remote means. By September 15, 1986, the licensee concluded that the valve could not be closed remotely, f
The licensee had three basic options:
(1) Derate to a very low power level, enter containment and open/ repair the valve (2) Shutdown (3) Request an amendment to their operating license to allow continued l operations. l l
The sta H finds that denial of the licensee's request for an amendment to their operating license would have placed the reactor in a shutdown or derated condition. Moreover, the licensee could not have foreseen this need for a license amendment in advance because it involved a component failure. Furthermore, even had the licensee requested an amendment of the license on the date the event occurred (September 8, 1986), the 14-day LC0 would have +/equired-NRC to issue the amendment on an emergency basis.
4.0 NO SIGNIFICANT HAZARDS CONSIDERATION The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amend-ment to an operating license for a facility involves no significant hazards considerations if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2)
Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
The licensee has provided an analysis of significant hazards considerations in its September 17, 1986 request which was amplified by their letter of September 19, 1985 for a license amendment. The licensee has concluded, with appropriate bases, that the proposed amendment meets the three standards in 10 CFR 50.92 and, therefore, involves no~significant hazards
- considerations. The Commission has also provided guidance concerning the i application of these standards by providing examples of- amendments i considered likely, and not likely, to involve a significant hazards consideration. These were published in the Federal Register on March 6,
'096 (5*. FR 7744). The NRC staff has made a preliminary review of the see's submittal. A discussion of these examples as they relate to aroposed amendment follows.
0: of the examples of actions involving no significant hazards consider-ation (vi) is a change which either may result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the standard review plan.
I
I i
The proposed amendment to the Technical Specifications may involve some increase, but in our judgment not a sionificant increase in the probability ,
or consequences of an accident previously evaluated because the potentially '
unisolated penetration is limited to a 3/a" line discharging to a closed system that returns to the containment or assuming a break in the RCIC steam line between valve M0v64 and MOV45,10CFp Part 100 limits will not be exceeded. '
This conclusion is based upon the fact that the fuel failure is not expected i because the make-up water from the fluid loss can be provided by the Feedwater i System or the High Pressure Core Spray system. The only activity available for release from the break is that which is present in the reactor coolant and steam lines prior to the break. Moreover, large break LOCA (Double-Ended Recirculation Line Break) with fuel failure assumes a closed loop in the RCIC system and filtration by the Standby Gas Treatment System for any valve packing leaks and meets the current accident analysis and 10CFR Part 100 limits.
The proposed amendment does not create.the possibility of a new or different kind of accident from any accident previously evaluated because line breaks such as may be postulated for the RCIC system have been considered in the analysis and design basis with many of the previous analyses bounding all the short-term consequences of the plant for this event. There may, however, be lonaer term environmental effects which would only impact the overall lifespan of the equipment and would be evaluated if such an event should occur. This change introduces no new mode of operation and no change in the plant design is being made.
The proposed amendment does not involve a significant reduction in the margin of safety because this event is bounded by a previous analysis and design basis. Therefore the margin of safety remains the same.
5.0 STATE CONSilLTATIONS In accordance with the Commission's regulations, consultation was held with officials of the State of Louisiana by telephone on September 19, 1986. The Manager, Nuclear Pro,iects of the Nuclear Energy Division, Office of Air Quality and Nuclear Energy, Department of Environmental Ouality of the State of Louisiana, had no objections to the proposed action.
6.0 PUBLIC COMMENTS None 0
7.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amour.ts, and no signifi-cant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has made a final no significant hazards consideration finding with respect to this amendment.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
8.0 CONCLUSION
The staff has concluded, based on the considerations discussed above that:
(1) the amendment does not (a) significantly increase tne probability or consecuences of an accident previously evaluated, (b) create the possibil-ity of a new or different kind of accident from any previously evaluated or (cl significantly reduce a safety margin and, therefore, the amendment does not involve a significant hazards consideration; (2) there is reason-able assurance that the health and safety of the public will not be endan-gered by operation in the proposed manner; and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the common defense and the security or to the health and safety of the public.
Principal Contributors: A. Notafrancesco C. Schulten S. Stern G. Thomas Dated: October 9,1986
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