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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 ML20236G8791987-10-23023 October 1987 Safety Evaluation Supporting Amend 86 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 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 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 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 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
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g UNITED STATES NUCLEAR REGULATORY COMMISSION r,, j WASHINGTON. D. C. 20555 k...../ SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 69 TO FACILITY OPERATING LICENSE NO. DPR-54 SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION i DOCKET NO. 50-312
1.0 INTRODUCTION
By letter dated December 17, 1984, as supplemented by letters dated March 14, 1985, and April 9, 1985, (Ref. 1) the Sacramento Municipal Utility District (SMUD or the licensee) submitted an application to modify the Technical Specifications for the Rancho Seco Nuclear Generating Station to permit operation for Cycle 7. A Cycle 7 reload report was also submitted with the above letters (Ref. 2). The fuel system, the nuclear, the themal-hydraulic and the accident and transient analyses of this reload are presented in the Reference 2 report and their evaluation follows. An evaluation of the.
proposed Technical Specification changes is also presented.
1.1 Description of the Cycle 7 Reload The Rancho Seco reload reactor core consists of 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods, 16 control rod guide tubes and one in-core instrument guide tube. The fuel pellets are dished on the ends and the cladding is Zircaloy-4. All fuel assemblies, including the 96 axial blanket assemblies and four axial blanket lead test assemblies, maintain a constant nominal fuel loading of 436.6 kg of uranium. There are seven different batches l
of fuel assemblies present in the Cycle 7 loading. Only 100 assemblies (40 of batch 88,56ofbatch9and4leadtestassembliesinbatch7)have.axialblankets.
The maximum initial enrichment of 3.43 w/o in U-235 is in batch There are 88.I 61 Ag-In-Cd full length control rnds and 56 burnable poison rod asse'rblies.
Finally, eight axial power shaping rods are provided for additional control of e . i.v N power distribistion.
8506210372 DR 850604 p ADOCK 05000312 PDR I - - , , , -n ,
- 2 The design of the Cycle 7 core was compared to the design of the Cycle 6 core which was used as the reference cycle.
2.0 FUEL SYSTEM DESIGN EVALUATION 2.1 Fuel Assembly Mechanical Design All fuel assemblies used in this reload are mechanically interchangeable into any core position. The fresh batch 9 (Mark B5Z) fuel has an axial blanket at the ends and incorporates Zircaloy spacer grids (instead of Inconel). The axial blanket design has been approved on the basis of the perfomance of lead test assemblies in batch 7 (Ref. 3). Zircaloy results in a better power
- distribution and better neutron economy compared to Inconel (Ref. 4). In addition,thebatch9holddownspringsaremadeofInconel718(insteadof Inconel 750) providing additional operating margin.
The Cycle 7 reload will include two regenerative neutron sources. For Mark B5 assemblies,the neutron source is built into the burnable poison rod assemblies.
The neutron source retainer design has been approved (References 5 and 6).
2.2 Cladding Stress, Strain and Collapse l .
, 7 The batch 6B power history has been shown to be the most limiting for cladding creep collapse for. Cycle 7. This case was compared with a generic analysis (which included increased peaking due to very low leakage) based on Reference 7.
l The analysis predicts a collapse time larger than 33,000 Effective Full Power Hours (EFPH)whiletheexpectedresidencetimeis29,213EFPH.
The cladding stress parameters for Cycle 7 are enveloped by a conservative analysis performed for Cycle 6, i.e., the reference cycle. The fuel pellet was designed
{ such that the design criteria of 1.0% plastic tensile strain is not to be exceeded for the worst local burnup and heat generation rate. The'gstrain analysis is also conservative based on the upper and lower tolerance values of the pellet and the cladding, respactively.
R v, V n r, the above restiltg the claMing design was found to be acceptable.
I 3
2.3 Fuel Thermal Design The Cycle 7 fuel analysis was perfonned using the approved TACO 2 code (Ref. 8).
The design of batch 9 (i.e., the new fuel in Cycle 7) is such as to be equivalent to the thermal design of the fuel batches present in Cycle 7 in the reniainder of the core. Densification effects were taken into account and the core protection limits were based on 20.4 kW/ft linear heat generation rate for centerline fuel melt.
The TAC 02 results indicate that the end of Cycle 7 maximum burnup will be (in
- batch 6B fuel) less than 43,000 MWD /MTU. The maximum burnup fuel internal pressure will be less than the nominal reactor coolant pressure of 2,200 psia.
The results, therefore, for the fuel thermal design are acceptable.
3.0 NUCLEAR DESIGN EVALUATION Physics characteristics for Cycle 7 have been calculated and compared to those corresponding to Cycle 6. The comparisons show that the physics parameters of the two cycles are similar., Analysis of the shutdown margin indicates that the minimum value is 2.57% 3k/k compared to the required shutdown margin of 1.0%
f 3k /k. The new assemblies in Cycle 7 (batch 9) contain an axial blanket, the design of which has been previously approved for Cycles 5 and 6. In addition, batch 9 fuel assemblies contain Zircaloy grids and "short-stack" lumped burn-able poisons and gray axial power shaping rods.
i Th physics arolysis was performed using the N0ODLE code which is a multidimensional, j two group reactor simulator (Ref. 9). The code has been compared to analytical results derived from PD007 and measured results and was shown to be of acceptable accuracy. The methodology of the NOODLE code has been recently aspr,oved by the NRC staff. I The axial power shaping rods will be withdrawn from the core during the last 30 EFPDs. Analysis has shown that during this period,the axial stability index l
. 4 will be negative, demonstrating that reactor operation will be stataa.
The results of the Cycle 7 reactor core physics analysis were found to be acceptable.
4.0 THERMAL-HYDRAULIC DESIGN EVALUATION In section 6.0 of Reference 2, the licensee described the thermal-hydraulic l
design for Cycle 7. Cycle 6 was used as the reference cycle for the thennal-hydraulic evaluation. Cycle 7 is a transition cycle to Mark BZ fuel assemblies
~
from Mark B. The difference of the two designs is that the Mark BZ assemblies
- (with Zircaloy grids) have increased fuel assembly pressere drop, thus diverting more flow to the Mark B assemblies. The operation of the Mark BZ assemblies has been reviewed and approved in Section 5 of Reference 4. A transition core penalty l
was applied in the Cycle 7 radial peaking factors. Analyses indicate that there is sufficient DNB margin to cover the transition penalty. No fuel rod bow penalty was included in the DNB as justified in Reference 10.
l, The important thermal-hydraulic parameters of Cycle 7 are the same as the corre-sponding parameters in Cycle 6. The methods used have been reviewed and approved.
> D dased on tnese results, we find the thermal-hydraulic design evaluation of l Cycle 7 acceptable.'
i l
5.0 ACCIDENT AND TRANSIENT ANALYSIS EVALUATION i
We have reviewed the information presented regarding the transient and accident analysis for Cycle 7. Comparison of the thermal-hydraulic parameters which detemine the outcome of the transients with those used in Cycle 6 indicates l
that the values are similar. In addition, comparison of the kineticparameters of Cycle 6 and the corresponding predicted values for Cycle 7 indicates that the Cycle 7 values are conservative with respect to their effect on trar,sient analyses.
Furthermore, a loss of coolant accident (LOCA) analysis was perfonned based on
. d ud criteria rene ted in Raf***rre 11. f.11 e'll .1 t i on'. t hew Da t
5 these criteria are conservative compared to those calculated for Cycle 7.
On the basis that these parameters were calculated by methods previously used and approved for Rancho Seco, we conclude that the transient and accident analyses have been treated properly and are acceptable.
6.0 EVALUATION OF TECHNICAL SPECIFICATIONS CHANGES We have reviewed the proposed changes to the following Technical Specifications for Cycle 7 operation:
2.1 Safety Limits, Reactor Core j 2.3 Limiting Safety System Settings Protective Instrumentation 3.2 High Pressure Injection and Chemical Addition Systems
. 3.5.2.5 Control Rod Positions 2.1 Safety' Limits, Reactor Core i
These modifications have been introduced because the BWC critical heat flux correlation was used in the calculation of the temperature - pressure limits.
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The BWC correlation has been approved. Therefore, we find the specified DNB SI'--~ limits acceptable to assure operation of the Rancho Seco Cycle 7 core within the applicable fuel' design and performance criteria.
2.3 Limiting Safety System Settings, Protective Instrumentation The modifications in this specification are a result of the BWC critical heat i
flux correlation and the DNB limits applied in the accident analysis. The
- BWC correlation has been approved. Therefore, we find the changes acceptable to assure operation of the Rancho Seco Cycle 7 core within the bounding values t
for the allowable LOCA peak linear heat rates. -
[
. .m..- - , - - , . . - -
6 3.2 High Pressure Injection and Chemical Addition Systems This modification change was necessary to maintain a minimum of 1% 4 k/k sub-
' critical reactivity for Cycle 7 under a cold shutdown condition with one stuck rod. The requirement is to increase the boric acid storage tank capacity fra 9,105 gallons to 9,586 gallons of 7,100 ppm of boron in solution. The capa' city of the present beric acid tank is 10,000 gallons. This modification has been derived using approved methods and the predicted Cycle 7 core characteristics are similar to those of previous core loadings.
3.5.2.5 Control Rod Positions The allowable power imbalance envelope for Cycle 7 has be6n computed using approved methods in such a manner that specified acceptable fuel design limits will not exceed the acceptance criteria. We find the Technical Specification change acceptable.
7.0 EVALUATION FINDINGS p We have reviewed the fuels, physics, thermal-hydraulic and accident analysis information presented in the Rancho Seco, Cycle 7 reload report as stated above. We find the proposed reload and the associated modified Technical Specifications acceptable.
8.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. We have determined that the amendment involves no significant increase N the amounts, and no significant change in the types, of any effluents th'at'ma,v be released offsite, and that there is no significant increase in individual or l com1ativa areno**innet -adiation evnnture. The Commission has previously l
l
7 issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public consnent on such finding.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set ferth 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.
9.0 CONCLUSION
We have concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will
. not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Comicsion's regulations and the issuance of this amendment will not be inimical to the comon defense and
! security or to the health and safety of the public. '
Dated: June 4, 1985 Principal Contributor: Lambros Lois N
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8 References
- 1. Letters: R. J. Rodriguez to Director, NRR, dated December 17, 1984, Serial: RJR 84-556, March 14,1985, Serial No. RJR 85-114 and April 9, 1985 Serial: RJR 85-169.
- 2. BAW-1850, " Rancho Seco Unit 1 Cycle 7 Reload Report". Volume 2, October 1984
- 3. BAW-1664, " Axial Blanket Lead Test Assembly", Licensing report, Babcock
. and Wilcox, Lynchburg, VA, September 1984.
i
- 4. BAW-1781P, " Mark BZ Fuel Assembly Design Report" Rancho Seco Cycle 7 Reload, Volume 1. Babcock and Wilcox, Lynchburg, VA, April 1983.
]
- 5. BAW-1496 "8PRA Retainer Design Report", Babcock and Wilcox, Lynchburg, VA, May 1978.
Il
- 6. Letter: J. F. Stolz to"J. H. Taylor (B&W) "BPRA Retainer Reinsertion".
- July 24, 1984 D 7. BAW-10084A, Rev. 2. " Program to Determine In-Reactor Performance of B&W Fuels; Cladding Creep-Collapse", Babcock and Wilcox, Lynchburg, VA, October 1978, i
i
- 9. BAW-10141PA, "fuci Pin Performance Analysis *, Y. H. Hsii, et.al., Babcock and Wilcox, Lynchburg, VA, June 1983.
- 9. BAW-10152, "NDODLE: A Multi-Dimensional, Two Group Reactor Simulat.or".
C. W. Nays, et al., Babcock and Wilcox, Lynchburg, VA, Septebmer I1984 4
(;
9
- 10. BAW-10147P-A, Rev. 1. " Fuel Rod Bowing in B&W Fuel Design", J. C. Moxley, et. al., Babcock and Wilcox, Lynchburg, VA, May 1983.
- 11. BAW-10103, Rev.1. "ECCS Analysis of B&W's 177-FA Lowered-Loop NSSS",
Babcock and Wilcox, Lynchburg, VA, September 1975.
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