Safety Evaluation Supporting Util 850524,1104 & 861105 Submittals Re Process Used to Identify Deviations from Emergency Response Guidelines.Justification Provided for Identifying Deviations AcceptableML20214W882 |
Person / Time |
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Site: |
Wolf Creek ![Wolf Creek Nuclear Operating Corporation icon.png](/w/images/4/42/Wolf_Creek_Nuclear_Operating_Corporation_icon.png) |
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Issue date: |
12/02/1986 |
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From: |
Office of Nuclear Reactor Regulation |
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To: |
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Shared Package |
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ML20214W869 |
List: |
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References |
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TAC-57368, NUDOCS 8612100444 |
Download: ML20214W882 (4) |
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20211N0161999-09-0202 September 1999 Safety Evaluation Supporting GL 95-07 to License NPF-42 ML20195D5261999-06-0202 June 1999 Safety Evaluation Approving Proposed ISI Program Alternative for Limited Reactor Vessel Shell Weld Exams & Relief Request from Requirements of ASME Code,Section XI ML20195H9801998-11-17017 November 1998 Safety Evaluation Supporting Proposed Changes to WCGS Radiological Emergency Response Plan ML20216J7791998-04-15015 April 1998 SER Approving Requests for Relief I1R-46 Through I1R-49 & I2R-21 Submitted by Licensee on 970523.Relief for Exam Category B-A,Item B1.12,RPV Shell Welds Deferred Until Licensee Satisfies Regulations for Augmented Rv Exam ML20216C2641998-04-0606 April 1998 SER Accepting Addl Info Re GL 92-08, Thermo- Lag 330-1 Fire Barriers, for Plant ML20217H3491998-03-31031 March 1998 SER Accepting Operational Quality Assurance Program Description Change for Wolf Creek Generating Station ML20217H7241998-03-30030 March 1998 SER Accepting Proposed Change to Operational Quality Assurance Program for Plant ML20202B9791997-11-26026 November 1997 Safety Evaluation Accepting Relief Requests 2VR-7 VR-8 from ASME Boiler & Pressure Vessel Code,Section XI Requirements for Inservice Testing Program ML20198Q2821997-10-24024 October 1997 Safety Evaluation Granting Relief for Second 10-yr Interval ISI Program Plan & Associated Requests for Plant ML20217J7071997-10-0909 October 1997 Safety Evaluation Accepting Second 10-yr ISI Interval Relief Request I2R-22 ML20141D0501997-06-23023 June 1997 Safety Evaluation Supporting Requests for Relief to Use ASME Code Case N-508-1 for Plant ML20147C1561997-01-31031 January 1997 Safety Evaluation Accepting Licensee Structural Integrity & Operability ML20247N1811989-09-20020 September 1989 Safety Evaluation Supporting Plant Inservice Testing Program & Request for Relief ML20236C7501989-03-0909 March 1989 SER Supporting Util 880801 Request for Relief from ASME Code Pump & Valve Inservice Testing Program Requirements for Bearing Temp for Safety Injection Pumps ML20153E6921988-09-0101 September 1988 Safety Evaluation Accepting Licensee Program for Equipment Classification & Vendor Interface ML20151H1411988-07-19019 July 1988 Safety Evaluation Accepting Util Response to Generic Ltr 83-28,Item 4.3 Re Reactor Trip Breaker Automatic Shunt Trip ML20151G5121988-07-15015 July 1988 Safety Evaluation Supporting Amend 17 to License NPF-42 ML20195K0801988-06-17017 June 1988 Safety Evaluation Supporting Amend 16 to License NPF-42 ML20155G1041988-06-0101 June 1988 Safety Evaluation Re Diesel Generator Circuit Breaker Manual Close Capability.Final Design Acceptable ML20153F8471988-05-0202 May 1988 Safety Evaluation Granting Util Relief from Section XI of ASME Code Re Replacement of Parts on Two Code Classed Components 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 ML20148J7351988-01-15015 January 1988 SER of First 10-yr Inservice Testing Program Plan & Requests for Relief from Requirements Concluding That Program as Modified by SER Provides Reasonable Assurance of Operational Readiness of safety-related Pumps & Valves ML20237D7591987-12-16016 December 1987 Safety Evaluation Concluding That Util Proposed ATWS Design in Compliance w/10CFR50.62 & Acceptable,Subj to Final Resolution of Tech Spec Issue & Completion of Certain Human Factors Engineering Reviews ML20236R3241987-11-12012 November 1987 Safety Evaluation Supporting Util Submittal of Rev 2 to First 10-yr Interval Inservice Insp Program Plan ML20205Q7131987-04-0101 April 1987 Safety Evaluation Re Util 861001 Submittal of Analysis of Large Break LOCA in Response to License Condition 2.C.(12). Analysis Acceptable & License Condition 2.C.(12) Satisfied ML20211P1991986-12-15015 December 1986 Safety Evaluation Supporting Util 860123 Response to PTS Rule.Matl Properties & Fast Neutron Fluence Calculations Acceptable Until End of Current License.Reevaluation of Rt (PTS) Required by 10CFR50.61 ML20214W8821986-12-0202 December 1986 Safety Evaluation Supporting Util 850524,1104 & 861105 Submittals Re Process Used to Identify Deviations from Emergency Response Guidelines.Justification Provided for Identifying Deviations Acceptable ML20215K9841986-10-22022 October 1986 SER Supporting Util 831115 & 860812 Responses to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Re post-maint Testing Requirements for safety-related Components,Based on Plans to Address Diesel Generator Testing Concerns ML20236N9221986-10-0707 October 1986 SER Accepting Util 840229 & 860529 Responses to Generic Ltr 83-28,Items 3.1.1 & 3.1.2 Re post-maint Testing of Reactor Trip Sys Components ML20215D8321986-10-0707 October 1986 SER Re Util 840229 & 860529 Responses to Generic Ltr 83-28, Items 3.2.1 & 3.2.2 Re post-maint Testing of safety-related components.Post-maint Testing Program Acceptable ML20215D6221986-10-0707 October 1986 SER Re Licensee Response to Generic Ltr 83-28,Item 4.5.1, Reactor Trip Sys Reliability (Sys Functional Testing). Reactor Trip Sys Functional Testing Acceptable ML20203G5361986-07-24024 July 1986 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 1.2 Re post-trip Review Data & Info Capability ML20211P9321986-07-21021 July 1986 Safety Evaluation Supporting Util Preventive Maint Program for Reactor Trip Breakers in Response to Generic Ltr 83-28 Items 4.1,4.2.1 & 4.2.2 ML20134F5061985-08-16016 August 1985 Safety Evaluation Supporting Util 850522 Description of Mods to Low Temp Overpressure Protection Sys of Rcs.Requirements of License Condition 2.C.(13) Met ML20134F5191985-08-16016 August 1985 Safety Evaluation Supporting Rev 12 to Facility Site Addendum Re Removal of 345 Kv Offsite Transmission Line to West Gardner Switching Station from Design ML20128Q8371985-07-0303 July 1985 SER Supporting Deletion of Pseudo Rod Cluster Control Assembly Drop Test at 50% Power.No Useful Info Would Be Gained from Performance of Test ML20128C8321985-06-26026 June 1985 SER Supporting Licensee Programs Re Generic Ltr 83-28,Item 1.1, Post-Trip Review 1999-09-02
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G1521999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Wolf Creek Generating Station.With ML20211N0161999-09-0202 September 1999 Safety Evaluation Supporting GL 95-07 to License NPF-42 ML20212A0251999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Wolf Creek Generating Station.With ML20217P6451999-08-30030 August 1999 Requests Commission Approval to Publish Encl Pr,Rg & SRP & to Issue Encl Ltr to Parties of Wolf Creek Transfer Proceeding Re Disposition of Existing Antitrust License Conditions in Event OL Transfer Approved ML20210R5741999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Wolf Creek Generating Station ML20210J1561999-07-29029 July 1999 Rev 0 to Wolf Creek Generating Station,Unit 1 Pressure & Temp Limts Rept ML20210R5921999-06-30030 June 1999 Revised Monthly Operating Rept for June 1999 for Wolf Creek Generating Station ML20209H0821999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Wolf Creek Generating Station ML20195D5261999-06-0202 June 1999 Safety Evaluation Approving Proposed ISI Program Alternative for Limited Reactor Vessel Shell Weld Exams & Relief Request from Requirements of ASME Code,Section XI ML20210R5871999-05-31031 May 1999 Revised Monthly Operating Rept for May 1999 for Wolf Creek Generating Station ML20195K1021999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Wolf Creek Generating Station ML20195K1071999-04-30030 April 1999 Revised MOR for Apr 1999 for Wolf Creek Generating Station ML20206P8261999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Wcgs.With ML20196L3401999-04-30030 April 1999 Rev 1 to WCGS Cycle 11 Colr ML20210R5841999-04-30030 April 1999 Revised Monthly Operating Rept for Apr 1999 for Wolf Creek Generating Station ML20205Q0761999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Wolf Creek Generating Station.With ML20207K5991999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Wolf Creek Generating Station.With ML20207K9761998-12-31031 December 1998 Annual SER 14,for Period 980101-1231, for WCGS ML20199E6531998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Wolf Creek Generating Station.With ML20195C0011998-12-31031 December 1998 Ks City Power & Light Co 1998 Annual Rept & Financial Statements as of 981231 & 1997 for Ks Electric Power Cooperative,Inc ML20195B9901998-12-31031 December 1998 Western Resources Annual Rept for 1998 ML20198D7321998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Wolf Creek Generating Station.With ML20195H9801998-11-17017 November 1998 Safety Evaluation Supporting Proposed Changes to WCGS Radiological Emergency Response Plan ML20195E7591998-11-10010 November 1998 WCNOC Proposed PASS Function Reduction ML20195D1791998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Wolf Creek Generating Station.With ML20154L4591998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Wolf Creek Generating Station.With ML20153G2771998-09-30030 September 1998 Rev 1 to WCAP-15079, Wolf Creek Heatup & Cooldown Limit Curves for Normal Operation ML20153G2851998-09-30030 September 1998 Rev 1 to WCAP-15080, Evaluation of Pressurized Thermal Shock for Wolf Creek ML20153G2691998-09-30030 September 1998 Rev 1 to WCAP-15078, Analysis of Capsule V from Wolf Creek Nuclear Operating Corp Wolf Creek Reactor Vessel Radiation Surveillance Program ML20153G7301998-09-23023 September 1998 Special Rept 98-003:on 980814,station Entered TS 3.3.3.6, Action Statment a Due to Inoperability of RVLIS B Train. Cause Has Not Yet Been Identified.Work Order 98-202813-000 Has Been Generated ML20151W1491998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Wcgs.With ML20237B7381998-08-14014 August 1998 Special Rept 98-001:on 980615,oxygen Analyzer on Wgs Was Declared Inoperable.Wgs Oxygen Analyzer OARC-1119A Was Indicating 0 Ppm on 980814 & Fluctuated Between 200 & 900 Ppm on 980615.Completed Work Order & Declared Wgs Operable ML20237B0841998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Wolf Creek Generating Station ML20236P3441998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Wolf Creek Generating Station ML20236P3481998-05-31031 May 1998 Corrected Page of MOR for May 1998 for Wolf Creek Generating Station ML20249A0171998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Wolf Creek Generating Station ML20249B2451998-05-18018 May 1998 Nonproprietary Version of Revised Chapters 4 & 5 to Rev 4 of HI-971769, Licensing Rept for Reracking of Callaway & Wolf Creek Nuclear Plants for Ue & Wcnoc. Chapters 4 & 5 Reflect Editorial Revs ML20248C3681998-05-18018 May 1998 Non-proprietary Version of Rev 4 to HI-971769, Licensing Rept for Reracking of Callaway & Wolf Creek Nuclear Plants ML20247H0901998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Wolf Creek Generating Station ML20216J7791998-04-15015 April 1998 SER Approving Requests for Relief I1R-46 Through I1R-49 & I2R-21 Submitted by Licensee on 970523.Relief for Exam Category B-A,Item B1.12,RPV Shell Welds Deferred Until Licensee Satisfies Regulations for Augmented Rv Exam ML20216C2641998-04-0606 April 1998 SER Accepting Addl Info Re GL 92-08, Thermo- Lag 330-1 Fire Barriers, for Plant ML20216F6101998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Wolf Creek Generating Station ML20217H3491998-03-31031 March 1998 SER Accepting Operational Quality Assurance Program Description Change for Wolf Creek Generating Station ML20217H7241998-03-30030 March 1998 SER Accepting Proposed Change to Operational Quality Assurance Program for Plant ML20216G1971998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Wolf Creek Nuclear Operating Corp ML20202H0721998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Wolf Creek Generating Station ML20217G4311997-12-31031 December 1997 Western Resources 1997 Annual Rept & Financial Statements for Years Ended 971231 & 1996 for Ks Electric Power Cooperative,Inc ML20203H0151997-12-31031 December 1997 Annual Operating Rept 13 for Jan-Dec 1997 ML20217G3711997-12-31031 December 1997 Kansas City Power & Light Co 1997 Annual Rept ML20216D7771997-12-31031 December 1997 Annual SER 12 for Jan-Dec 1997, for Wolf Creek Generating Station 1999-09-30
[Table view] |
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!" n NUCLEAR REGULATORY COMMISSION O :p WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION VANSAS GAS AND ELECTRIC C0l1PANY KANSAS CIT,Y POWER AND LIGHT C0f'PANY KANSAS ELECTRIC POWER COOPERATIVE, INC.
WOLF CREEK GENERATING STATION DOCKET NO. 50 482 BACKGROUND The Safety Evaluation Report related to the Operation of Wolf Creek Generating Station (NUREG-0881, Supplement 5) published in March 1985 pointed out that Kansas Gas and Electric Company (KG&E) needed to identify the deviations from the generic Emergency Response Guidelines (ERGS) that exist in the Wolf Creek Ertergency Procedures. On August 30, 1985 the staff sent KG&E a request with regard to certain specific, potentially safety significant deviations and re-quested KG&E to review its methods for identifying technical deviations from
_ the ERGS.
KG&E's responses of November 4, 1985 and November 5, 1986 provided adequate justification for the specific deviations identified by the staff. They also described how KGAE identified potentially safety-significant deviations from the ERGS and listed the deviations together with an explanation of each deviation.
We have reviewed all of the deviations from the ERGS identified by the licensee and by the staff and have found the deviations to be acceptable as discussec below.
1 EVALUATION The deviations from the ERGS identified by the staff were discovered as a result of the Detailed Control Room Design Review (DCRDR). On May 24, 19PF the licensee reported its findings resulting from the verification of the l task analysis for the SNUPPS DCRDR. In reviewing these findings, the NRC staff questioned whether certain of these findings with regard to control room design wculd also necessitate deviations frera the ERGS in order to provide the reactor operator with adequate guidance for ccping with emergencies. The staff questioned findings 1, 6, 8, 9 and 10 discussed below:
Finding 1: Boron Injection Tank (BIT) Ten.perature 1
ERG ECA 0.0 Loss of All AC Power, Step 13a(2) calls for BIT temperature I indication but none is nrnvided at the Wolf Creek Generating Station (hCGS).
8612100444 861202 2 PDR ADOCK 0500 P
v, 9
This ERG step calls for a check of BIT temperature and is based on the Westing-house ERG reference plert that raaintains the boron concentration in the BIT at 21,000 ppr... The bli is maintained at an elevated temperature to keep the boror in solution. At the WCGS, the boron cercentration has been reduced from 21,000 to 2,000 ppm. At the lower concentration, all of the boron will remain in solution at ar.ibient temperatures. Therefore the cperator does not need to know the BIT temperature arid should not be concerned with it in this Lriegency Orerating Procedure (EOP).
The licensee has deleted the UCGS-EOF step that required the ocerater to check BIT temperature. This is ecceptable to the staff.
Finding 6:
WCGS Emergency Procedure E-1 (corresponding to ERG Procedure E-1: Loss of Reactor Or Secondary Ceolant), step 13A, required operator action at greater than 535 GPl; (Resioual Heat Removal water flow to the RCS). Since control room indicators are graduated in increments of 100 GPM, the value of 535 GPi' cannot be read accurately.
The licensee has reevaluated this matter and found that the value 500 GPM, as read on the flow indicators in the control room, provides sufficient indication of PPR flow into the RCS. This WCGS Emergency Procedure was changed to require operator action after verifying RPR ficw to the RCS to be greater than 500 GPM. We find this change to be acceptable.
Findings 8 and 9:
Findings eight and nine alleged that no instrumentation was provided in the control room for the verification of flow of Component Cooling Water (CCW) to the RHR heat exchanger and to the (Reactor Coolant Purrp) Seal Water Heat
. Exchanger. -
The licensee reviewed this matter and found nurrerous ways of verifying CCW flow including the Balance of Plant Cemruter Display in the control room, service loop flow indicators ard beat exchanger temperature indications. In addition, the control room annunciators signal conditions of high flow and low flow in the CCW System and the reactor coolant pump Seal Return Heat Exchanger.
Cased cr tFis information, we find that this llCCS instrumentation is adequatc for implementation of the generic Festinghouse ERGS and therefore re deviction from the ERGS is necessary. The direct application of the pertirient LPGs is therefore appropriate at the VCCS.
Finding 10:
Finding 10 asserts that the generic EEG L-5, Steam Generator Tube Rupture, Step P, calls for steamline radiation tronitors t.hereas none are provided in the WCGS.
~'
g The purpose of this EPG Step is to identify which stean, generator may have a leaking tott. The operator observes stean generator water level and indications of radioactivity in the secondary system. The k' CGS E0Ps direct the operator to utilize steam generator water level, the turbine driven
, euxiliary feedwater pump steam eybaust radiation monitor, the steant generator power operated relief valve p1tre radiation monitors, the steen generator blowcowr. radiation monitors, and steam cererator sampling to identify the steam generator with the leaking tube. Based on the availability of these alternative neans for identifying the leakino steem generator by the use of radiation monitors, we have concluded that these UCGS instrumentation and procedures will provide the reactor operators with adequate inferration and guidance for identifying the leaking steam generator.
As a result of its review of the ERGS and background material for potentially safety significant deviations from the ERGS, the licensee identified additional deviations that required further censideration.
Five of these deviations from the ERGS are based on differences in the design of the kCGS from that of the Westinghouse ERG refererce plant. The design differences are:
- 1. WGS has the added feature of automatic closure of the pressurizer PORY block valve at "2185 psig oecreasing". Thus, where the ERGS call for verification of operability and preliminary opening of one block valyc, the WCGS E0Ps call for verification of operability, but a block valve n.ay or may not remain open and the block valves would be opened intnediately before opening a PCRV.
- 2. Circuitry has been added at WCGS to ensure that undervoltaoe (UV) relays do not actuate when a reactor coolant pump (RCP) is started. The circuitry delays operation of the UV relays and is activated by resetting
. the Safety injection (SI) Actuation System. Thus a step was added in the procedure for natural circulation cooldown to reset SI prior to starting an RCP.
- 3. The UCGS does not have steam line radiction monitors as in the reference plant, but it does have stear generator PORV plume monitors that are not in the reference plant. Appropriate devietions from the ERGS were incorporated into the PCCS procedures as discusseo under finding 10 above.
4 It the LC65, the control rod position indicators, including the red bottom lights, are rot cperational during a loss of all AC power.
Operator instructions to check the rod bottcr lights and the rod position indicators have not been put into the KCCS E0P for loss of All AC Power. The operator mil observe neutron flux and the position of reactor trip and bypass breukers to verify reactor trip.
- 5. At the WCGS, in order to mininize ECC5 flow in the event of loss of en,ergency coolant recirculotten, certain valves may be coerated from the control rocm to modulate or throttle the flew c1 emergency coolant,
4 4
l 1 /! -
whereas in the reference plantothe valves are either fully open or fully closed. Since modulation or throttling is the preferable method of controlling flow, the WCGS E0P instructs the operator to throttle flow rather than fully open or close valves and turn purros on or off.
We have reviewed these five deviations that are baseo on design differences between the WCGS and the ERG reference plant. We have concluded that these are safety significant deviations from the ERGS that are necessary to achieve the cbjectives of the ERGS. Furthennore, the procedural changes adopted et the WCGS provide appropriate instructions to the operator to cope with the pertinent emergencies.
The licensee's procedures also deviate from the ERGS involving reactor trip or safety injection. The ERGS have a guideline, E-0, for Reactor Trip or Safety Injection (SI) and a separate guideline ES-0.1 for Reactor Trip Response (without SI) whereas at the WCGS there is a procedure EMG-0 for Safety Injection and a procedure EPG-ES-02 for Reactor Trip. The licensee has taken the first four steps of E-0, that is the verification or check of Reactor Trip, Turbine Trip, AC Power and SI actuation and placed them in both EMG-0 and ENG-ES-02. Thus the WCGS operator can start with either the Reactor Trip Guideline or the Safety Injection Guideline and then transfer to the one which is more appropriate. The ERG Guidelines require the operator to start with E-0 and then, if appropriate, transfer to ES-0.1, Reactor Trip.
The licensee has indicated that simulator exercises have demonstrated the benefits of giving the operator direct access to the Reactor Trip Procedure EMG-ES-02. lle have reviewed the licensee's description of this deviation from the ERGS and found that the licensee has properly accommodated the transitions involved and that the KCGS procedures will accomplish the objectives of the corresponding ERGS. Therefore, this deviation from the ERGS is acceptable to the staff.
The licensee also provided additional guidance in two areas that was not provided explicitly in the ERGS. Additional guidance was pruvided throughout the EOFs in the event that the reactor vessel liquid inventory instrurrentation is not operational. Also, additional precautions are taken in EPG C-II, Un-contrclied Depressurization of All Stecm Generators, that will provide further assurance that cold overpressurization of the reactor vessel will not occur.
These additions to the E0Ps do not depart from the strategy cr irtent of the ERGS and do rot irhibit the implementation of other operator actions. These additions are therefore acceptable to the staff.
CONCLUSION We have reviewed the dif ferences between the 1.'estinghouse Owners' Group ERGS and the Knif Creek E0Ps that were identified from the staff's examination of the DCRDR and the licensee's revieu to identify deviations from the ERGS. Based on this review, we have concluded that these procedural changes adopteo at the WCGS provide adequate guidance and information to the operator to cope with emergencies and achieve the pertinent objectives of the bestinghouse generic ERGS.
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