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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20211H7921999-08-13013 August 1999 Safety Evaluation Supporting Amend 126 to License DPR-54 ML20198G8271997-08-22022 August 1997 Safety Evaluation Supporting Amend 125 to License DPR-54 ML20059G9621990-09-10010 September 1990 Safety Evaluation Supporting Amend 115 to License DPR-54 ML20247B3611989-07-17017 July 1989 Safety Evaluation Supporting Amend 112 to License DPR-54 ML20245E1161989-06-20020 June 1989 Safety Evaluation Supporting Amend 111 to License DPR-54 ML20245A1991989-06-0909 June 1989 Safety Evaluation Supporting Amend 110 to License DPR-54 ML20248B6321989-06-0505 June 1989 Safety Evaluation Supporting Amend 108 to License DPR-54 ML20248B9361989-06-0505 June 1989 Safety Evaluation Supporting Amend 107 to License DPR-54 ML20248B9621989-06-0505 June 1989 Safety Evaluation Supporting Amend 109 to License DPR-54 ML20247P1761989-05-30030 May 1989 Safety Evaluation Accepting Generic Ltr 83-28,Item 4.5.2 Re on-line Testing of Reactor Trip Sys ML20247M9231989-05-23023 May 1989 Safety Evaluation Supporting Amend 106 to License DPR-54 ML20247K7261989-05-23023 May 1989 Safety Evaluation Supporting Amend 105 to License DPR-54 ML20247K6871989-05-16016 May 1989 Safety Evaluation Supporting Amend 104 to License DPR-54 ML20246F4671989-05-0404 May 1989 Safety Evaluation Re Inservice Testing Program & Requests for Relief Re ASME Class 1,2 & 3 Pumps & Valves.Program Acceptable ML20245F9041989-04-18018 April 1989 Safety Evaluation Supporting Amend 103 to License DPR-54 ML20248E9371989-03-29029 March 1989 Safety Evaluation Supporting Amend 102 to License DPR-54 ML20155D3131988-09-28028 September 1988 Safety Evaluation Supporting Amend 100 to License DPR-54 ML20151L5511988-07-14014 July 1988 Redistributed Safety Evaluation Supporting Amend 99 to License DPR-54 ML20151A2771988-07-13013 July 1988 SER Supporting Util Actions to Prevent Failure of Ammonia Tanks Which May Result in Incapacitation of Control Room & Technical Support Ctr Personnel ML20195C5571988-06-0808 June 1988 SER Supporting Util Responses to Generic Ltr 83-28,Item 2.1 (Part 1) Re Equipment Classification (Reactor Trip Sys Components) ML20195C9311988-06-0808 June 1988 SER Accepting Util Response to Generic Ltr 83-28,Item 2.1 (Part 2) Re Vendor Interface Programs (Reactor Trip Sys) ML20150F2851988-03-28028 March 1988 Safety Evaluation Supporting Resolution of Tdi Diesel Engine Vibration Problems at Facility ML20148J8611988-03-17017 March 1988 Safety Evaluation Supporting Amend 98 to License DPR-54 ML20153B2911988-03-15015 March 1988 Safety Evaluation Supporting Amend 97 to License DPR-54 ML20055E3041988-02-12012 February 1988 Safety Evaluation Supporting Util 871223 & 880111 Proposed Changes to Tech Specs,Including Reducing Lower Limits of Detection for Liquid Radioactive Effluents ML20149L6961988-02-12012 February 1988 Safety Evaluation Supporting Amend 95 to License DPR-54 ML20149L2761988-02-0909 February 1988 Safety Evaluation Supporting Amend 94 to License DPR-54 ML20148C7321988-01-0505 January 1988 Safety Evaluation Supporting Amend 93 to License DPR-54 ML20237B4141987-12-0707 December 1987 Safety Evaluation Supporting Amend 92 to License DPR-54 ML20236X1771987-12-0303 December 1987 Safety Evaluation Supporting Amend 91 to License DPR-54 ML20236X1111987-11-13013 November 1987 Safety Evaluation Supporting Amend 90 to License DPR-54 ML20236Q2461987-11-10010 November 1987 Safety Evaluation Supporting Util & Related Submittals Re Design Mods to Emergency Electrical Distribution Sys (Ref Tdi Diesel Generators) ML20236J2671987-11-0303 November 1987 Safety Evaluation Supporting Amend 89 to License DPR-54 ML20245C7051987-10-27027 October 1987 Safety Evaluation Supporting Amend 87 to License DPR-54 ML20245C7831987-10-27027 October 1987 Safety Evaluation Supporting Amend 88 to License DPR-54 ML20236D1361987-10-23023 October 1987 Safety Evaluation Supporting Util 870826 Request to Use Repair & Replacement Program Contained in ASME Section XI 1983 Edition Including Addenda Through Summer 1983 ML20236G8791987-10-23023 October 1987 Safety Evaluation Supporting Amend 86 to License DPR-54 ML20238D4151987-09-0303 September 1987 Evaluation of Engineering Rept ERPT-E0220 Re Reactor Regulation of Util Approach to Compliance W/Reg Guide 1.75 for New Diesel Generator Installation at Plant.Licensee Approach to Demonstrating Compliance Acceptable ML20238B1771987-08-27027 August 1987 Safety Evaluation Supporting Amend 85 to License DPR-54 ML20236E3731987-07-24024 July 1987 Safety Evaluation Supporting Existing & Proposed Mods to Meteorological Program & W/Planned Improvements,Facility Will Satisfy Min Meteorological Emergency Preparedness Requirements of 10CFR50.47 & 10CFR50,Apps E & F ML20205T3021987-03-31031 March 1987 Safety Evaluation Supporting Amend 84 to License DPR-54 ML20205R7871987-03-26026 March 1987 Safety Evaluation Concluding Util 831104 Response to Generic Ltr 83-28,Item 4.5.2, Reactor Trip Sys Reliability On-Line Testing Permits on-line Functional Testing of Sys,Including Diverse Trip Features of Reactor Trip Breakers ML20205C3951987-03-13013 March 1987 Safety Evaluation Re Fracture Toughness Requirements for Protection Against PTS Events (10CFR50.61).Util 860123 Submittal Re Matl Properties & Fast Neutron Fluence of Reactor Vessel Acceptable ML20206B6561987-03-13013 March 1987 Corrected SER Re Fracture Toughness Requirements for Protection Against PTS Events (10CFR50.61).Util 860123 Submittal Re Matl Properties & Fast Neutron Fluence of Reactor Vessel Acceptable ML20205J0971987-03-11011 March 1987 Safety Evaluation Re Sys Selected for Facility Sys Review & Test Program.Sys Constitutes Adequate Scope for Sys Review & Test Program ML20211P2601987-02-19019 February 1987 Safety Evaluation Accepting Util 860116 Request for Amend to License DPR-54,redefining Fire Area Boundaries Required to Be Operable to Separate safety-related Fire Areas & Reassessing Adequacy of Components in Fire Area Assemblies ML20210D1011987-02-0303 February 1987 Safety Evaluation Supporting Issuance of Amend 83 to License DPR-54 ML20215N6881986-11-0404 November 1986 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 4.4 Re Improvements in Maint & Test Procedures for B&W Plants ML20212N4351986-08-13013 August 1986 Safety Evaluation Supporting Amend 82 to License DPR-54 ML20206F9071986-05-19019 May 1986 Safety Evaluation Supporting Util Request for Exemption from Requirements of 10CFR50,App R,Subsection Iii.L Re Capability to Achieve Cold Shutdown in 72 H 1999-08-13
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20211H7921999-08-13013 August 1999 Safety Evaluation Supporting Amend 126 to License DPR-54 ML20195D1901999-05-0606 May 1999 Annual Rept ML20195H8571998-12-31031 December 1998 1998 Annual Rept for Smud. with ML20155D4801998-10-27027 October 1998 Amend 3 to Rancho Seco DSAR, Representing Updated Licensing Basis for Operation of Permanently Shutdown & Defueled Rancho Seco Nuclear Facility During Permanently Defueled Mode ML20248C4301998-05-0606 May 1998 Annual Rept, Covering Period 970507-980506 ML20249A7831997-12-31031 December 1997 1997 Smud Annual Rept ML20198G8271997-08-22022 August 1997 Safety Evaluation Supporting Amend 125 to License DPR-54 ML20217D3271997-07-30030 July 1997 Update of 1995 Decommissioning Evaluation for Rancho Seco Nuclear Generating Station ML20140A6371997-05-0606 May 1997 Annual Rept, Covering Period 960507-970506 ML20140G4481997-05-0101 May 1997 Part 21 Rept Re Potential Defect in Component of Dsrv & Dsr Enterprise Standby Diesel Generator Sys.Recommends That Springs Be Inspected on Periodic Basis,Such as During Refueling Outages ML20137W8151997-03-20020 March 1997 Amend 1 to Post Shutdown Decommissioning Activities Rept ML20141J2711996-12-31031 December 1996 Smud 1996 Annual Rept ML20138L1231996-11-13013 November 1996 Smud Rancho Seco Incremental Decommissioning Action Plan, Rev 0,961113 ML20129E7151996-10-14014 October 1996 Defueled SAR for Rancho Seco ML20059H6821994-01-17017 January 1994 Revised Rancho Seco Quality Manual ML20058K3841993-12-0909 December 1993 Part 21 Rept Re Potential Defect in Component of Dsrv & Dsr Enterprise Standby DG Sys,Regarding Potential Problem W/ Subcover Assembled Atop Power Head ML20056E5171993-08-31031 August 1993 Technical Review Rept, Tardy Licensee Actions ML20059K1981993-05-0606 May 1993 Annual Rept, Covering Period from 920501- 930506,consisting of Shutdown Statistics,Narrative Summary of Shutdown Experience & Tabulations of Facility Changes, Tests & Experiments,Per 10CFR50.59(b) ML20128C9641993-02-0202 February 1993 Informs Commission of Status of Open Issues & Progress of Specified Facilities Toward Decommissioning ML20127H2301993-01-15015 January 1993 Part 21 Rept Re Potential Defeat in Component of Dsrv & Dsr Enterprise Standby DG Sys.Starting Air Distributor Housing Assemblies Installed as Replacement Parts at Listed Sites ML20126B0421992-12-17017 December 1992 Final Part 21 Rept Re Potential Problem W/Steel Cylinder Heads.Initially Reported on 921125.Caused by Inadequate Cast Wall Thickness at 3/4-inch-10 Bolt Hole.Stud at Location Indicated on Encl Sketch Should Be Removed ML20125C7161992-12-0707 December 1992 Part 21 Rept Re Possibility for Malfunction of Declutching Mechanisms in SMB/SB-000 & SMB/SB/SBD-00 Actuators. Malfunction Only Occurs During Seismic Event.Balanced Levers May Be Purchased from Vendor.List of Affected Utils Encl ML20127P5861992-11-23023 November 1992 Followup to 921005 Part 21 Rept Re Potential Defect in SB/SBD-1 Housing Cover Screws.Procedure Re Replacement of SBD-1 Spring Cover Bolts Encl.All Fasteners Should Be Loosened & Removed.List of Affected Utils Encl ML20126E6771992-08-0303 August 1992 Rev 7 to Rancho Seco Quality Manual NL-90-451, Monthly Operating Rept for Oct 1990 for Rancho Seco Nuclear Generating Station1990-10-31031 October 1990 Monthly Operating Rept for Oct 1990 for Rancho Seco Nuclear Generating Station ML17348B5061990-10-0909 October 1990 Part 21 Rept Re Zener Diode VR2 on Power Supply Board 9 1682 00 106 Possibly Being Installed Backwards ML20059G9621990-09-10010 September 1990 Safety Evaluation Supporting Amend 115 to License DPR-54 NL-90-443, Monthly Operating Rept for Aug 1990 for Rancho Seco1990-08-31031 August 1990 Monthly Operating Rept for Aug 1990 for Rancho Seco ML20217A5711990-08-28028 August 1990 Final Engineering Rept,Assessment of Spent Fuel Pool Liner Leakage NL-90-439, Monthly Operating Rept for Jul 1990 for Rancho Seco Nuclear Generating Station1990-07-31031 July 1990 Monthly Operating Rept for Jul 1990 for Rancho Seco Nuclear Generating Station ML20055F8591990-07-16016 July 1990 Special Rept 90-11:on 900613,06,25,18,21 & 28,fire Barriers Breached More than 7 Days & Not Made Operable in 14 Days. Corrective Actions:Operability of Fire Detectors Verified on One Side of Breached Barriers NL-90-423, Monthly Operating Rept for June 1990 for Rancho Seco Nuclear Generating Station1990-06-30030 June 1990 Monthly Operating Rept for June 1990 for Rancho Seco Nuclear Generating Station ML20055C6301990-05-21021 May 1990 Special Rept 90-09:on 900424,25,30,31 & 0502,fire Barriers Inoperable for More than 7 Days,Per Tech Spec 3.14.6.2 Requirement.Hourly Fire Watches Established & Penetrations & Doors Returned to Operable Status ML20055C6291990-05-21021 May 1990 Special Rept 90-08:on 900419,fire Pump Batteries Inoperable When Surveillance Procedure SP.206 Not Performed by Due Date.Caused by Test Frequency Incorrectly Changed from Weekly to Monthly.Surveillance Schedule Revised ML20058B6521990-05-0404 May 1990 Rev 0 to ERPT-M0216, Property Loss Study for Rancho Seco Nuclear Generating Station in Long Term Defueled Mode 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 NL-89-634, Monthly Operating Rept for Aug 1989 for Rancho Seco Nuclear Generating Station1989-08-31031 August 1989 Monthly Operating Rept for Aug 1989 for Rancho Seco Nuclear Generating Station NL-89-598, Monthly Operating Rept for Jul 1989 for Rancho Seco Nuclear Generating Station1989-07-31031 July 1989 Monthly Operating Rept for Jul 1989 for Rancho Seco Nuclear Generating Station ML20247B3611989-07-17017 July 1989 Safety Evaluation Supporting Amend 112 to License DPR-54 NL-89-556, Monthly Operating Rept for June 1989 for Rancho Seco Nuclear Station1989-06-30030 June 1989 Monthly Operating Rept for June 1989 for Rancho Seco Nuclear Station ML20245E1161989-06-20020 June 1989 Safety Evaluation Supporting Amend 111 to License DPR-54 ML20245B6651989-06-15015 June 1989 Part 21 Rept 150 Re Potential Defect in Component of Dsr Standby Diesel Generator.Cause of Failure Determined to Be Combination of Insufficient Lubrication to Bushings.Listed Course of Action Recommended at Next Scheduled Engine Maint ML20245A1991989-06-0909 June 1989 Safety Evaluation Supporting Amend 110 to License DPR-54 ML20248B9361989-06-0505 June 1989 Safety Evaluation Supporting Amend 107 to License DPR-54 ML20248B6321989-06-0505 June 1989 Safety Evaluation Supporting Amend 108 to License DPR-54 ML20248B9621989-06-0505 June 1989 Safety Evaluation Supporting Amend 109 to License DPR-54 NL-89-517, Monthly Operating Rept for May 1989 for Rancho Seco Nuclear Generating Station1989-05-31031 May 1989 Monthly Operating Rept for May 1989 for Rancho Seco Nuclear Generating Station ML20247P1761989-05-30030 May 1989 Safety Evaluation Accepting Generic Ltr 83-28,Item 4.5.2 Re on-line Testing of Reactor Trip Sys ML20247N7491989-05-30030 May 1989 Special Rept 89-13:on 890328-29,specific Activity of Primary Sys Exceeded Limits in Administrative Procedure & Chemistry Control Commitment.Possibly Caused by Power Reduction.Isotopic Analysis Continued ML20247K7261989-05-23023 May 1989 Safety Evaluation Supporting Amend 105 to License DPR-54 1999-08-13
[Table view] |
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SAFETY EVALUATION BY THE.0FFICE.0F. NUCLEAR REACTOR REGULATION SUPPORTING. AMENDMENT NO.112TO. FACILITY.0PERATING LICENSE.DPR 54 3ANCH0.SEC0 NUCLEAR. GENERATING. STATION,. UNIT.1 DOCKET.NO. 50-312
1.0 INTRODUCTION
By letters dated February 28, 1986, May 14, 1987, and August 31, 1988, the Sacramento Municipal Utility District, the licensee for the Rancho Seco Nuclear Generating Station, proposed Technical Specification (TS) changes to Appendix A of Operating License DPR-54 for Rancho Seco. The proposed changes were requested to revise the Rancho Seco TS Table 3.6-1, " Safety Features Containment Isolation Valves" to add 12 new valves into the table and to increase the valve closure time to 25 seconds for all valves listed in the table, including new valves.
The current TS Table 3.7-1 lists 7 safety features containment isolation valves with required maximum valve closure times ranging from 3 to 22 seconds. The proposed amendment adds 12 valves to the table, each with a 25-second closure time, and increases the valve closure time to 25 seconds for 23 selected valves.
The remaining 12 valves (out of 35 valves currently listed in the table) have not changed the maximum allowable closure time in the proposed amendment. All of these safety features isolation valves offer a direct pathway from the containment to the environment.
2.0 EVALUATION In this evaluation, the staff grouped 35 isolation valves (23 selected valves from the current table plus 12 newly added valves) as follows:
- 1. Reactor Building Purge Valves (4 valves) 2 .. Valves with no automatic isolation features (remote-manual operation) and normally in closed position (10 valves)
- 3. Valves with automatic isolation features and normally in either closed or open position (21 valves)
E .1 Reactor Buildino. Inlet.and Outlet. Purge Valves.(SFV-53503, SFV-53504, 3FV-53604, and.5FV-53605)
These four reactor building inlet and outlet purge valves are administratively locked closed. The Rancho Seco TS Section 3.6.7 states that the purge valves should be closed with their respective breakers de-energized, except during cold shutdow or refueling. It further states that the valve position should be veri'ied at leas; monthly. Therefore, the valve closing time for these purge valves are only relevant to a postulated fuel handling accident during refuelint op ration.
8907240l01 890717 hDR ADOCK 05000312 i PDC
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The staff reevaluated the offsite radiological consequences due to the. increased ostulated fuel handling accident.
valve In the Rancho closureSeco timeSafety (25 seconds)
Evaluationfollowing Report a p(SER) dated June 1973, the staff previously evaluated a postulated fuel handling accident using assumptions stated in Section 15.3 of the SER. In this evaluation, the staff calculated i the incremental offsite doses attributable to the delayed valve closure time j (25 seconds) using the same assumptions used in the previous analysis. The SER does.not state the valve closure time used previously and therefore, the staff assumed that the zero valve closure time was used.
The staff's calculated offsite doses resulting from 25 second valve closure time following a postulated fuel handling accident inside containment are presented in the attached Table 1 along with previously calculated offsite doses due to the accident. As shown in the table, the potential overall offsite doses for a postulated fuel handling accident including incremental doses attributable to the delayed valve closure time of 25 seconds are still within the acceptance criteria specified in the Standard Review Plan (SRP)
Section'15.7.4. Therefore, the staff finds that the proposed valve closure time of 25 seconds is acceptable.
2.2 Valves With No Automatic. Isolation. Features.(Remote Manual Operation).and Normally in Closed. Position (10. Valves)
- 1. HV-53617 Reactor Building Hydrogen Purge Line
- 2. HV-53618 Reactor Building Hydrogen Purge Line
- 3. HV-70040 Hydrogen Monitoring Isolation Line
- 4. HV-70041 Hydrogen Monitoring Isolation Line
- 5. HV-70042 Hydrogen Monitoring Isolation Line
- 6. HV-70043 Hydroaen Monitoring Iso 16: ion Line
- 7. HV-70044 Hydrogen Monitoring Isolation Line
- 8. HV-70045 Hycrogen Monitoring Isolation Line 1
- 9. HV-70046 Hydrogen Monitoring Isolation Line I
- 10. HV-70047 Hydrogen Monitoring Isolation Line The above valves are normally closed, remotely-operated one-inch diameter hand valves. The operation of these valves is administratively controlled An operator is in attendance whenever these valves are operated for the purpose of sampling and surveillance. Otherwise, the valves remain in a closed position. The licensee stated in the proposed amendment that these /alves were installed in 1983 as a part of the Post Accident Sampling System.
Therefore, the potential offsite doses are not relevant to and not affected by
l the increased valve closure time (25 seconds) of these normally closed valves.
The valve closure times are only specified and added to the table since they are qualified and designated as containment isolation valves. Therefore, the staff finds that the increased closure times for these valves are acceptable.
2.3 All Other. Remaining Valves (21. Valves)
These valves are normally in either the closed or open position with automatic isolation features.
In the proposed amendment, the licensee modeled, for the purpose of estimatin5 the mass flow rate expected from the containment to the environment following the receipt of a LOCA signal, all 10 containment penetrations (one 6" RC System Drain, two 4" RS Sump Drain and PC Pump Seal Water Return, one 3" RC System Vent, one 21/2" RS System Letdc.m, four 1" sample lines, and one 3/4" sample line) with their 19 isolation valves (except one 12-inch reactor building pressure equalizer line with two isolation valves) as one large penetration (10-inch) having an equivalent cross sectional area. The licensee estimated average mass flow rates from time zero to 25 seconds to be approximately 2.4 x 104 and 1.8 x 104 ft3 through the 12-inch reactor building pressure equalizer line and one 10-inch modeled line respectively. The staff used the LOCA blowdown mass release rate of 4.9 x 105 lbs for the first 25 seconds as given in the Rancho Seco USAR Table 14.4-4.
The Rancho Seco USAR Figures 1.4.2-31 and 14.2-32 show peak fuel cladding temperature (hot spot) as a function of time after a LOCA for an 8.55 ft2 double-ended break in cold leg pipe at the reactor feed pump discharge. The licensee stated that this break resulted in the highest calculated fuel cladding temperature. According to these figures, the cladding temperature of approximately 2000 F. will be reached at 25 seconds after a LOCA. A maximum hot spot cladding temperature of 2299'F. at 38.5 seconds was calculated by the licensee meeting the NRC Interim Criteria of 2300 F. used in 197? (this limit has since been lowered by 10 CFR 50.46 to the current valve of 2200 F.).
Some of the fuel rods may be expected to experience claddir.g perforation-deformation failure due to the heatup transient (fuel-cladaing temperature excursion) during the first 25 seconds after a LOCA. Therefore, the staff used, in accordance with SRP 6.2.4, an iodine spike of 60 mci /yn (dose equiva-lent iodine-131) in the LOCA blowdown as source term. The source term activity will be immediately available for release through the containment isolation '
valve pathways (one 12" reactor building equilizer line and one 10" modeled line opening) for the first 25 seconds after the rer.eipt of a LOCA signal. No credit was given for removal of fission products in the staff's analyses (containment spray, charcoal absorbers, iodine par';ition, iodine plate-out, etc.).
The staff's calculated offsite doses resulting from 25 second valve closure time following a LOCA are also presented in Table 1 along with previously calculated offsite doses due to the LOCA as shown in the SER Table 15.0. As shown in Table 1, the potential overall offsite doses for a LOCA including mm_--__.__m_._m_m
1 1
-4 incresc9al doses attributable to the delayed valve closure time of 25 seconds are still within the dose reference values specified in 10 CFR Part 100.
Therefore, the staff finds that the proposed valve closure time of 25 seconds is acceptable.
3.0
SUMMARY
Based on our review, we find that calculated offsite doses including incremental doses attributable to the increased valve closure time are still within the exposure reference guidelines of 10 CFR Part 100 and are within the acceptance criteria given in Standard Review Plan Section 15.7.4. Therefore, we find the proposed changes to the Rancho Seco TS concerning the safety features contain-ment isolation valve closure time are acceptable.
4.0 CONTACT.WITH. STATE.0FFICIAL The NRC staff has advised the Chief of the Radiological Health Branch, State Department of Health Services, State of California, of the proposed determination of no significant hazards consideration. No comments were received.
5.0 ENVIRONMENTAL CONSIDERATION
This amendment involves changes in the installation or use of a f acility com-ponent 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 amounts, and no significant 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 previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 61.22(c)(9). Pursuantto10CFR51.22(b),noenvironmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
6.0 CONCLUSION
We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safet of the public will not be endangered by operation in the proposed manner, 2 (y) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to common defense and security or to the health and safety of the public.
Principal Contributor: J. Lee i
Dated: July 17, 1989 l l
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i Table 1- ,
Potential Offsite' Doses-Due'to Desian Basis' Accidents Exc1'usion Area- -Lov Populat' ion-Accidents Boundary - Zone-
.(rem) _ (rem)-
Thyroid -Whole Body. ' Thyroid.. Whole BodyL
. Fuel Handling 2 34 4.- 2'-
<1 Purge Valves 2 <
.1 - <1 '<1
- Total .. 36 <5 <3 <2' SRP Criteria Limit 75 6= '75 '6 Y
loss of-Coolant 1 267- '.8 48 'l Delayed Valve' Closures - 22 <1 3 - <1.'
Total. 289 <9 51 .<2-10 CFR 100 Limit 300 25 300- 25 2SER Tab'le 15.'0 225 second valve closure time
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