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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20211B0271999-08-13013 August 1999 SER Accepting Second 10-year Interval Inservice Insp Requests for Relief RR-A16,RR-A17 & RR-B9 for Plant, Unit 1 ML20195K2871999-06-16016 June 1999 Safety Evaluation Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20199H5931999-01-20020 January 1999 Safety Evaluation Accepting Thermo-Lag Re Ampacity Derating Issues for Plant ML20155B6781998-10-28028 October 1998 Safety Evaluation Accepting Proposed Reduction in Commitment Changes in QA Program Matl Receipt Insp Process ML20236R1441998-07-15015 July 1998 SER Related to Quality Assurance Program Description Changes for Davis-Besse Nuclear Power Station,Unit 1 ML20236K3981998-06-30030 June 1998 SER Accepting in Part & Denying in Part Relief Requests from Some of ASME Section XI Requirements as Endorsed by 10CFR50.55a for Containment Insp for Davis-Besse Nuclear Power Station,Unit 1 ML20236K5131998-06-29029 June 1998 Safety Evaluation Accepting Proposed Alternate Emergency Operations Facility Location for Davis-Besse Nuclear Power Station,Unit 1 ML20198R4771998-01-13013 January 1998 SER Approving Second 10-year Interval Inservice Inspection Program Plan Requests for Relief for Davis-Besse Nuclear Power Station,Unit 1 ML20128L3001996-10-0202 October 1996 SER Supporting Dbnp IPE Process of Identifying Most Likely Severe Accidents & Severe Accident Vulnerabilities ML20058M9591993-09-28028 September 1993 SE Accepting Licensee Response to GL 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implication of Control Sys in LWR Nuclear Power Plants.' ML20057A3791993-08-20020 August 1993 SE Concluding That Second 10-yr Interval Inservice Insp Program Plan for Plant Has Unacceptable Exam Sample as Discussed in Encl Inel TER ML20056G4301993-08-18018 August 1993 Safety Evaluation Re Inservice Testing Program Requests for Relief.Licensee Made Changes to Subj Program to Include Exercising & fail-safe Testing of Auxiliary Feedwater Valves AF-6451 & AF-6452,in Response to TER Anomaly 8 ML20056B2721990-08-20020 August 1990 Safety Evaluation Granting Relief from ASME Code Repair Requirements for ASME Code 3 Piping ML20248D8271989-09-29029 September 1989 Safety Evaluation Accepting Util 890228 & 0630 Submittals Presenting Proposed Designs to Comply w/10CFR50.62 ATWS Rule Requirements ML20247E6901989-09-0505 September 1989 Safety Evaluation of Audit of Facility Design for Resolution of IE Bulletin 79-27 Re Loss of non-Class IE Instrumentation & Control Power Sys Bus During Operation.Preventive Maint & Testing Program Should Be Developed for Bus Power Sources ML20247J8731989-05-18018 May 1989 Safety Evaluation Supporting Amend 133 to License NPF-3 NUREG-0660, Safety Evaluation Accepting Util 840301 & 870420 Responses to NUREG-0737,Item 1.C.1,except Where Noted in Section 21989-05-0303 May 1989 Safety Evaluation Accepting Util 840301 & 870420 Responses to NUREG-0737,Item 1.C.1,except Where Noted in Section 2 ML20196D9601988-12-0808 December 1988 Safety Evaluation Re Util Response Concerning Auxiliary Feedwater Sys Reliability Study.Util Should Ensure That Sys Mods Do Not Result in Net Reduction in Sys Reliability ML20207K7911988-10-0404 October 1988 Safety Evaluation Supporting Operation in Cycle 6 W/O Removing Flaws in Cracked HPI Nozzle ML20148D0391988-01-19019 January 1988 SER Accepting Util 831209 Response to Generic Ltr 83-28,Item 4.5.2 Re Reactor Trip reliability-on-line Testing.Unit Designed to Permit on-line Functional Testing of Diverse Trip Features of Reactor Trip Breakers ML20147C2631988-01-12012 January 1988 Safety Evaluation Accepting Util Responses to Generic Ltr 83-28,Item 2.1 (Part 1) Re Equipment Classification of Reactor Trip Sys Components ML20149F9621988-01-11011 January 1988 SER Accepting License Response to Generic Ltr 83-28,Item 2.2.1 Re Equipment Classification Programs for safety- Related Components ML20236U6531987-11-27027 November 1987 Safety Evaluation Re Effects of Errors in Util Analysis of Small Break Loca.Use of Incorrect Values in Analysis Results in Incomplete Compliance w/10CFR50,App K.Plant Poses No Risk to Public Health Due to Meeting 10CFR50.46 Requirements ML20236T3871987-11-25025 November 1987 SER Re Conformance to Reg Guide 1.97 Concerning post-accident Monitoring Instrumentation.Design Acceptable ML20211G8641987-02-11011 February 1987 Safety Evaluation Supporting Licensee Responses to Generic Ltr 83-28,Item 3.2.3, Post-Maint Testing & All Other Safety-Related Components ML20207Q6711987-01-0909 January 1987 Safety Evaluation Supporting Util Re Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events ML20212C9231986-08-0404 August 1986 Sser Supporting Util Reanalyses of 74 Masonry Walls ML20212G6021986-08-0404 August 1986 Safety Evaluation Accepting Util Response to NRC Request for Addl Info Re Question 29 Concerning Training for Infrequent, Critical & Difficult Tasks ML20212R3421986-06-10010 June 1986 SER Re Progress Made by Licensee in Area of Plant Maint.New Maint Organization Functioning W/No Major Weaknesses.Region III Will Continue to Monitor Maint at Plant to Assure Continued Control & Progress ML20206F3041986-05-27027 May 1986 Safety Evaluation Accepting Util Review Re Cold Pressurization of Secondary Side of Steam Generator During Testing on 850906 ML20203N9141986-04-17017 April 1986 Safety Evaluation Accepting Proposed Mods to Safety Actuation Sys Re Shared Power Supply Returns as Result of 801205 Actuation NUREG-0103, Safety Evaluation Supporting Relief from Inservice Insp Requirements of ASME Boiler & Pressure Vessel Code Requiring Quarterly Stroke Testing of RCS Valves & Denying Relief from Code Requirements for Valve RC-11 Testing1986-03-0606 March 1986 Safety Evaluation Supporting Relief from Inservice Insp Requirements of ASME Boiler & Pressure Vessel Code Requiring Quarterly Stroke Testing of RCS Valves & Denying Relief from Code Requirements for Valve RC-11 Testing ML20137J6891986-01-0808 January 1986 Safety Evaluation Supporting Util Identification of Root Causes for Spurious 850609 Steam & Feedwater Line Rupture Control Sys Low Level Actuation,Closure of MSIVs & Actions Taken to Prevent Recurrence.Plant Restart Acceptable ML20137J7031986-01-0808 January 1986 SER Supporting Util Identification of Root Causes for Source Range Nuclear Instrumentation Channels Inoperability During 850609 Steam & Feedwater Line Rupture Control Sys Low Level Actuation & Corrective Actions Taken.Plant Restart Approved ML20136B7841985-12-24024 December 1985 SER Re Util Development of Systematic & Thorough Troubleshooting Plans to Investigate 850609 Incident Leaving Redundant source-range Nuclear Instrumentation Channels Inoperable.Restart & Power Operation Now Acceptable ML20137R1611985-11-12012 November 1985 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Items 4.2.1 & 4.2.2 Re Preventive Maint & Trending of Parameters ML20138P0681985-10-30030 October 1985 Safety Evaluation Accepting Licensee 831107 & 850709 Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,4.1 & 4.5.1 Re post-maint Testing & Reactor Trip Sys Reliability (vendor-related Mods & Sys Functional Test Description) ML20138P6391985-10-30030 October 1985 SER Re Proposed Mods to Maint Program.Licensee Should Identify Actions to Be Completed Prior to or After Restart. Region Should Actively Monitor Licensee Progress ML20133Q0441985-10-24024 October 1985 Safety Evaluation Supporting Util 831107 Response to Generic Ltr 83-28,Item 3.1.3 Re post-maint Testing (Reactor Trip Sys Components).Eg&G Technical Evaluation Rept Encl ML20062E0871982-07-29029 July 1982 Safety Evaluation Supporting Licensee Request for Relief from ASME Boiler & Pressure Vessel Code Section XI Hydrostatic Test (Insp) Requirements ML20214J2801979-12-20020 December 1979 Safety Evaluation Re Preliminary Design for Upgrading Present control-grade Anticipatory Reactor Trip Sys for Loss of Main Feedwater & Turbine Trip to safety-grade ML20125B1331968-12-16016 December 1968 Safety Evaluation Re Piqua Nuclear Power Facility Retirement.T Hamrick to J Schlesinger Encl 1999-08-13
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217K1231999-10-14014 October 1999 Revised Positions for DBNPS & Pnpp QA Program ML20217D5441999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Davis-Besse Nuclear Power Station.With ML20211R0811999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Davis-Besse Nuclear Power Station,Unit 1.With ML20211B0271999-08-13013 August 1999 SER Accepting Second 10-year Interval Inservice Insp Requests for Relief RR-A16,RR-A17 & RR-B9 for Plant, Unit 1 ML20210Q8541999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Davis-Besse Nuclear Power Station,Unit 1.With ML20209E6231999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Davis-Besse Nuclear Power Station,Unit 1.With ML20195K2871999-06-16016 June 1999 Safety Evaluation Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20195F4871999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Davis-Besse Nuclear Power Station,Unit 1.With ML20207E8011999-05-19019 May 1999 Non-proprietary Rev 2 to HI-981933, Design & Licensing Rept DBNPS Unit 1 Cask Pit Rack Installation Project ML20207F4351999-05-0404 May 1999 Rev 1 to DBNPS Emergency Preparedness Evaluated Exercise Manual 990504 ML20206M6341999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Davis-Besse Nuclear Station,Unit 1.With ML20205M2931999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Davis-Besse Nuclear Power Station.With ML20207J1461999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Davis-Besse Nuclear Power Station,Unit 1.With ML20199H5931999-01-20020 January 1999 Safety Evaluation Accepting Thermo-Lag Re Ampacity Derating Issues for Plant ML20204J6751998-12-31031 December 1998 1998 Annual Rept for Dbnps,Unit 1,PNPP,Unit 1 & BVPS Units 1 & 2 ML20199E2501998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Davis-Besse Nuclear Power Station,Unit 1.With ML20206B0101998-12-31031 December 1998 1998 Annual Rept for Firstenergy Corp, for Perry Nuclear Power Plant & Davis-Besse Nuclear Power Station.Form 10-K Annual Rept to Us Securities & Exchange Commission for Fiscal Yr Ending 981231,encl ML20205K5781998-12-31031 December 1998 Waterhammer Phenomena in Containment Air Cooler Swss ML20197J3441998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Davis-Besse Nuclear Power Station,Unit 1.With ML20195D0001998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Davis-Besse Nuclear Power Station,Unit 1.With ML20155B6781998-10-28028 October 1998 Safety Evaluation Accepting Proposed Reduction in Commitment Changes in QA Program Matl Receipt Insp Process ML20154H5801998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Davis-Besse Nuclear Power Station,Unit 1.With ML20151W1611998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Dbnps.With ML20237E3171998-08-21021 August 1998 ISI Summary Rept of Eleventh Refueling Outage Activities for Davis-Besse Nuclear Power Station ML20237B1681998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Davis-Besse Nuclear Power Station,Unit 1 ML20236U5011998-07-23023 July 1998 Special Rept:On 980624,Unit 1 Site Damaged by Tornado & High Winds.Alert Declared by DBNPS Staff,Dbnps Emergency Response Facilities Activiated & Special Insp Team Deployed to Site by Nrc,As Result of Event ML20236R1441998-07-15015 July 1998 SER Related to Quality Assurance Program Description Changes for Davis-Besse Nuclear Power Station,Unit 1 ML20236N7451998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Davis-Besse Nuclear Power Station,Unit 1 ML20236K3981998-06-30030 June 1998 SER Accepting in Part & Denying in Part Relief Requests from Some of ASME Section XI Requirements as Endorsed by 10CFR50.55a for Containment Insp for Davis-Besse Nuclear Power Station,Unit 1 ML20236K5131998-06-29029 June 1998 Safety Evaluation Accepting Proposed Alternate Emergency Operations Facility Location for Davis-Besse Nuclear Power Station,Unit 1 ML20248F7441998-05-31031 May 1998 Reactor Vessel Working Group,Response to RAI Regarding Reactor Pressure Vessel Integrity ML20249A4121998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Davis-Besse Nuclear Power Station,Unit 1 ML20196B5221998-05-23023 May 1998 10CFR50.59 Summary Rept of Facility Changes,Tests & Experiments Dbnps,Unit 1 for 960602-980523 ML20236E7581998-05-19019 May 1998 Rev 0 to Davis-Besse Unit 1 Cycle 12 Colr ML20236N7501998-04-30030 April 1998 Rev 2 to Monthly Operating Rept for Apr 1998 for Davis-Besse Nuclear Power Station,Unit ML20247F6721998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Davis-Bess Nuclear Power Station,Unit 1 ML20249A4141998-04-30030 April 1998 Revised Monthly Operating Rept for Apr 1998 for Davis-Besse Nuclear Power Station,Unit 1 ML20217P8041998-04-0707 April 1998 11RFO OTSG ECT Insp Scope ML20216B4041998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Davis-Besse Nuclear Power Station,Unit 1 ML20216C5131998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Davis-Besse Nuclear Power Station,Unit 1 ML20202D3721998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Davis-Besse Nuclear Power Station,Unit 1 ML20199G6321998-01-26026 January 1998 Rev 1 to Davis-Besse Unit 1,Cycle 11,COLR ML20198R4771998-01-13013 January 1998 SER Approving Second 10-year Interval Inservice Inspection Program Plan Requests for Relief for Davis-Besse Nuclear Power Station,Unit 1 ML20198K7931997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Davis-Besse Nuclear Power Station,Unit 1 ML20217K6401997-12-31031 December 1997 1997 Annual Rept First Energy ML20203A3931997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for Davis-Besse Nuclear Power Plant,Unit 1 ML20198S5371997-10-31031 October 1997 Monthly Operating Rept for Oct 1997 for Davis-Besse Nuclear Power Station ML20217H7701997-09-30030 September 1997 Monthly Operating Rept for Sept 1997 for Davis-Besse Nuclear Power Station,Unit 1 ML20216H3261997-08-31031 August 1997 Monthly Operating Rept for August 1997 for DBNPS ML20217K0241997-07-31031 July 1997 Monthly Operating Rept for Jul 1997 for Davis Besse Nuclear Power Station,Unit 1 1999-09-30
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STAFF EVALUATION REPORT INDIVIDUAL PLANT EXAMINATION DAVIS-SESSE NUCLEAR POWER STATION. UNIT NO. 1 DOCKET NO. 50-346 1.0 INTIRBUCTION On February 26, 1993, the Toledo Edison Company submitted the Davis-Besse Nuclear Power Station (DBNPS) Individual Plant Evaluation (IPE) in response to Generic Letter 88-20, " Individual Plant Examination For Severe Accident Vulnerabilities," and associated supplements. On June 22, 1995, the staff sent questions to the licensee requesting additional information. The licensee responded in a letter dated September 11, 1995.
A " Step I" review of the DBNPS IPE submittal was performed a1d involved the efforts of Brookhaven National Laboratory (BNL). The Step 1 review focused on whether the licensee's method was capable of identifying vulnerabilities.
Therefore, the review considered -(1) the <:ompleteness of the information and (2) the reasonableness of the results given the DBNPS design, operation, and history. A more detailed review, a " Step 2" review, was not performed for this IPE submittal. The details of BNL's review findinas are included in the technical evaluation report-(TER) appendix to this staff evaluation report (SER).
In accordance with GL 88-20, Toledo Edison Company proposed to resolve Ui) resolved Safety Issue (USI) A-45, " Shutdown Decay Heat Removal Requirements." The licensee also proposed, and the staff agreed to consider USI A-17. " Systems Interactions in Nuclear Power Plants," for resolution with the submission of the internal flood portion of the IPE submittal.
In addition, the following Generic Safety Issues (GSIs) were included by the licensee in the IPE submittal for resolution:
GSI-23, " Reactor Coolant Pump Seal Failures,"
GSI-105, " Interfacing Systems Loss-of-Coolant-Accidents in Pressurized Water Ret.ctors,"
GSI-128, " Electric Power Reliability and Related Issues,"
GSI-143, " Availability of Chilled Water Systems and Room Coolers," and GSI-153, " Loss of Essential Service Water in Light Water Reactors." i Section II of this SER discusses resolution of these USIs and GSIs.
l 9610150004 961002 '
PDR ADOCK 05000346 '
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2.0 EVALUATION Davis-Besse is a Babcock and Wilcox (B&W) pressurized water reactor (PWR) with a large, dry containment. The reactor coolant system (RCS) is a " raised loop design" with the steam generators above the reactor core to allow an inventory of RCS coolant to flow back into the core in the event of a LOCA and to promote natural circulation. This raised loop design of the RCS is unique in that all other B&W plants are lower loop designs. The DBNPS IPE has estimated a core damage frequency (CDF) of 6.6E-05 per reactor-year from internally initiated events, including the contribution from internal floods. The DBNPS CDF compares reasonably with that of other B&W PWR plants. The CDF contribution from station blackout (SB0) is less than that typical of many PWRs (about 2 percent) due to the existence of two turbine-driven auxiliary feedwater (TDAFW) pumps, and one motor-driven auxiliary feedrater (MDAFW) pump, with the TDAFW pumps capable of manual control after a DC power failure.
The SB0 CDF is further reduced by the existence of an SB0 diesel generator, sturdy reactor coolant pump seals, and a backup supply of emergency feedwater.
Transients contribute 86.4 percent (including 2 percent from SBO), LOCAs 8.6 percent, internal flooding 3 percent, interfacing system LOCA 1.3 percent, and steam generator tube rupture 0.7 percent. The important system / equipment contributors to the estimated CDF that appear in the top sequences are TDAFW pump failures, component cooling water (CCW) system failures, and failure of makeup /high pressure injection (HPI) cooling. The licensee's Level 1 hnalysis appears to have examined the significant initiating events and dominant accident sequences.
Based on the licensee's IPE process used to search for decay heat removal (DHR) vulnerabilities, and review of the DBNPS plant-specific features, the staff finds the licensee's DHR evaluation consistent with the intent of the USI A-45, Decay Heat Removal Reliability, and is, therefore, acceptable.
Furthermore, the licensee did not identify any vulnerabilities with respect to GSIs 23, 77, 105, 128, 143, and 153. According to GL 88-20, if a licensee ;
concludes "that no vulnerability exists at its plant that is topically '
associated with any USI or GSI, the staff will consider the USI or GSI resolved for a plant upon review and acceptance of the results of the IPE."
Accordingly, the staff concludes that the licensee resolved USIs A-45 and A-17 and GSI-77, 105, 128, 143, and 153. Regarding GSI-23, the Commission has decided not to take additional rulemaking action at this time and plans to issue a generic communication on this issue at a future date; therefore, the staff cannot conclude that GSI-23 has been resolved.
The licensee performed a human reliability analysis (HRA) to document and quantify potential failures in human-system interactions and to quantify human-initiated recovery of failure events. The licensee identified the following operator actions as important in the estimate of the CDF: failure to start the MDAFW pump and initiate makeup /HPI cooling, failure to locally control the TDAFW pumps during a SBO, and failure to start the SB0 diesel generator and control the TDAFW pumps.
i The licensee evaluated and quantified the results of the severe accident progression through the use of a containment event tree and considered I
_. a s. ., -- o. - ,- - . - - - - - -x .. - . - . - - - +
uncertainties in containment response through the use of sensitivity analyses.
The licensee's back-end analysis appeared to have considered important severe accident phenomena. According to the licensee, the DBNPS conditional containment failure probabilities are as follows: early containment failure is 0.6 percent with direct attack of debris on the containment side wall being the primary contributor; late containment failure is 9.1 percent with containment overpressurization and basemat melt-through being the primary contributors, and bypass is 2.6 percent with interfacing system LOCA and SGTR being the primary contributors. According to the licensee, the containment remains intact 87.7 percent of the time. Isolation failure (due to failure to isolate the sump drain line), estimated at 1.4 percent of CDF, was lumped into
" intact" due to the small radiological releases associated with it.
Radiological releases, not categorized into early and late components, are dominated by transients including SB0. The licensee's response to containment performance improvement (CPI) program recommendations is consistent with the intent of GL 88-20 and associated Supplement 3.
According to the licensee, some insights and unique plant safety features identified at DBNPS are:
- 1. The turbine-driven main feedwater pumps will continue to run for most transients, as the pump flow output is automatically matched to the decay heat level.
- 2. The two TDAFW pumps can be manually controlled locally in SB0 conditions, i even after depletion of the batteries. However, with the usual system ,
configuration, failure to control one pump will lead to failure of both l TDAFW pumps due to water carryover in the steam lines.
{
- 3. The MDAFW pump has to be started manually by the operators. If offsite power is lost, the pump is powered from the SB0 diesel generator only.
- 4. One pressurizer power operated relief valve (PORV) and two safety valves can be used for makeup /HPI ccoling, i.e., feed and bleed. This gives Davis-Besse a diversity of options for makeup /HPI cooling.
- 5. Three emergency diesel generators are available for onsite AC power.
- 6. There is a high level of service water (SW) system and CCW system redundancy with each system having three 100% capacity pumps. In addition, the dilution pump can be used as a backup SW system pump.
- 7. A reasonably large area is available under the reactor vessel for corium spreading. This results in a corium thickness of about 10 inches, at nominal corium density, if the entire available corium mass is spread in the reactor cavity.
- 8. The containment shell is protected by a 1.5 foot wide by 2.5 foot high !
curb at the basemat floor elevation. This feature protects the steel shell from direct contact with corium if it was to relocate to the lower containment and, thus, serves to lower the probability of this shell failure mode, ,
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! The licensee defined a vulnerability as either (1) a CDF significantly higher
? then 1.0E-4 per reactor year with one or a few aspects of the plant design or ;
j operating practices contributing to such a high frequency, (2) a single plant '
- feature (or a few features) which causes a disproportionately high t contribution to the CDF, or (3) a CDF that is very sensitive to a highly j i uncertain aspect of plant response. Based on this definition, the licensee !
did not identify any vulnerabilities. i l
j Plant improvements, however, were identified. These improvements, listed l below, have been implemented ,
- 1. Shedding of DC loads. At the time of the IPE analysis, procedural
, guidance existed only for the case when power was unavailable to both AC .
! divisions. The E0P procedural revisions provide guidance when only one I j division of AC power is lost. l
- 2. Enhanced procedures and training for makeup to the boric water storage ;
- tank (BWST) in the SGTR scenarios where the BWST is depleted by injection i
- before the RCS is depressurized.
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- 3. Improved fuel oil monitoring for the station blackout diesel generator 1 j (SBODG). The SBODG has a limited fuel oil supply (4 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of run l 4 time). Operational procedures have been revised to include direction for j I monitoring the level and consumption rate of fuel oil during emergency ;
i operations. Specific direction is provided to initiate refill efforts for l j the supply tank upon reaching a predetermined level. ;
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3.0 CONCLUSION
j Based on the above findings, the staff notes that (1) the IPE is complete with i regard to the information requested by GL 88-20 (and associated guidance in l NUREG-1235), and (2) the IPE results are reasonable given the DBNPS design,
- operation, and history. As a result, the staff concludes that DBNPS's IPE l process is capable of identifying the most likely severe accidents and severe accident vulnerabilities, and therefore, that the DBNPS IPE has met the intent 1 of GL 88-20.
l It should be noted that the staff's review primarily focused on the licensee's
! ability to examine DBNPS for severe accident vulnerabilities. Although
! certain aspects of the IPE were explored in more detail than others, the review is not intended to validate the accuracy of the licensee's detailed findings (or quantification estimates) that stemmed from the examination.
- Therefore, the SER does not constitute NRC approval or endorsement.of any IPE
- j material for purposes other than those associated with meeting the intent of ;
GL 88-20, i
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APfud 4X ,
t TECHNICAL EVALUATION REPORT )
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FOR THE 1 DAVIS-BESSE NUCLEAR POWER STATION l
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