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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217B5401999-10-0606 October 1999 Safety Evaluation Supporting Amend 193 to License DPR-40 ML20211J9321999-09-0202 September 1999 Safety Evaluation Concluding That Licensee Proposed Alternatives Provide Acceptable Level of Quality & Safety. Proposed Alternatives Authorized for Remainder of Third ten- Yr ISI Interval for Fort Calhoun Station,Unit 1 ML20210G2181999-07-27027 July 1999 Safety Evaluation Supporting Amend 192 to License DPR-40 ML20210D9951999-07-22022 July 1999 Safety Evaluation Supporting Amend 191 to License DPR-40 ML20206L4241999-05-10010 May 1999 Safety Evaluation Supporting Corrective Actions to Ensure That Valves Are Capable of Performing Intended Safety Functions & OPPD Adequately Addressed Requested Actions Discussed in GL 95-07 ML20206M2601999-05-0606 May 1999 SER Concluding That Licensee IPEEE Complete Re Info Requested by Suppl 4 to GL 88-20 & IPEEE Results Reasonable Given FCS Design,Operation & History ML20205Q5831999-04-15015 April 1999 Safety Evaluation Supporting Amend 190 to License DPR-40 ML20198S3771998-12-31031 December 1998 Safety Evaluation Supporting Amend 189 to License DPR-40 ML20198S4831998-12-31031 December 1998 Safety Evaluation Supporting Amend 188 to License DPR-40 ML20154M4881998-10-19019 October 1998 Safety Evaluation Supporting Amend 186 to License DPR-40 ML20154N2411998-10-19019 October 1998 Safety Evaluation Supporting Amend 187 to License DPR-40 ML20236V4891998-07-30030 July 1998 Safety Evaluation Relating to Response to GL 87-02,suppl 1 for Fort Calhoun Station,Unit 1 ML20248C0671998-05-21021 May 1998 Safety Evaluation Granting Licensee Request for Exemption from Technical Requirements of 10CFR50,App R, Fire Protection Program for Nuclear Power Facilities Operating Prior to 790101 ML20217L7201998-03-23023 March 1998 Safety Evaluation Supporting Amend 185 to License DPR-40 ML20203M4161998-02-0303 February 1998 Safety Evaluation Supporting Amend 184 to License DPR-40 ML20203A4291998-01-26026 January 1998 Safety Evaluation Supporting Amend 183 to License DPR-40 ML20199L0711997-11-24024 November 1997 Safety Evaluation Supporting Amend 182 to License DPR-40 ML20198Q4031997-10-28028 October 1997 Safety Evaluation Re Control Room Habitability Requirements ML20137L6241997-03-27027 March 1997 Safety Evaluation Supporting Amend 181 to License DPR-40 ML20134N7751997-02-13013 February 1997 Safety Evaluation Supporting Amend 180 to License DPR-40 ML20134M6171997-02-13013 February 1997 Safety Evaluation Denying Licensee Request for Approval to Use ASME Section XI Code Case N-416-1 W/Proposed Exception & Code Case N-498-2 as Alternative to Required Hydrostatic Pressure Test ML20133P9161997-01-23023 January 1997 Safety Evaluation Accepting Revised Temperature Limits for DG-1 & DG-2 ML20133C2771996-12-30030 December 1996 Safety Evaluation Supporting Amend 179 to License DPR-40 ML20132F4911996-12-0909 December 1996 Safety Evaluation Related to Individual Plant Evaluation Omaha Power District,Fort Calhoun Station,Unit 1 ML20134M0871996-11-19019 November 1996 Safety Evaluation Supporting Request for Relief from Modifying Supports SIH-3,SIS-63,SIS-65 & RCH-13 at Fort Calhoun Station ML20129H3371996-10-25025 October 1996 Safety Evaluation Supporting Amend 178 to License DPR-40 ML20128F6441996-10-0202 October 1996 Safety Evaluation Supporting Amend 177 to License DPR-40 ML20129G3131996-09-27027 September 1996 Safety Evaluation Supporting Amend 176 to License DPR-40 ML20059J1831994-01-14014 January 1994 Safety Evaluation Supporting Amend 160 to License DPR-40 ML20059J2491994-01-14014 January 1994 Safety Evaluation Supporting Amend 159 to License DPR-40 ML20058G9371993-12-0303 December 1993 Safety Evaluation Supporting Amend 158 to License DPR-40 ML20058F5951993-11-22022 November 1993 Safety Evaluation Supporting Amend 157 to License DPR-40 ML20058C7491993-11-18018 November 1993 Safety Evaluation,Authorizing Alternative,On One Time Basis Only,W/Conditions That Licensee Perform Volumetric Exam of nozzle-to-vessel Welds During First Refueling Outage of Third 10-yr Insp Interval ML20059L7081993-11-10010 November 1993 Safety Evaluation Accepting Licensee Proposed Changes to Low Power Physics Testing Program ML20059G6601993-10-29029 October 1993 Safety Evaluation Supporting Amend 156 to License DPR-40 ML20057E3471993-10-0101 October 1993 Safety Evaluation Advising That Based on Determination That Alternative Testing Consistent w/OM-10,paragraph 4.3.2.2. Requirements,No Relief Required ML20056E5411993-08-12012 August 1993 Safety Evaluation Supporting Amend 155 to License DPR-40 ML20056E5371993-08-10010 August 1993 Safety Evaluation Supporting Amend 154 to License DPR-40 ML20056D6801993-07-26026 July 1993 Safety Evaluation Supporting Amend 153 to License DPR-40 ML20128B8241993-01-26026 January 1993 Safety Evaluation Supporting Amend 149 to License DPR-40 ML20128D4511992-11-30030 November 1992 Safety Evaluation Accepting Evaluation of 120-day Response to Suppl 1 to GL 87-02, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors, Unresolved Safety Issue A-46 ML20062G6621990-11-19019 November 1990 Safety Evaluation Supporting Amend 134 to License DPR-40 ML20216K0661990-11-14014 November 1990 Safety Evaluation Denying Util 900221 & 0622 Requests for Exemption from App R of 10CFR50 for Fire Area 34B,upper Electrical Penetration Room.Current Level of Fire Protection Does Not Meet Section III.G.2 Requirements ML20062B6161990-10-12012 October 1990 Safety Evaluation Supporting Amend 133 to License DPR-40 ML20055G0221990-07-0606 July 1990 Safety Evaluation Supporting Amend 132 to License DPR-40 ML20246A0741989-08-17017 August 1989 Safety Evaluation Re Inservice Testing Program for Pumps & Valves ML20245H9031989-08-15015 August 1989 Safety Evaluation Re Generic Ltr 83-28,Item 2.1 (Part 1), Equipment Classification (Reactor Trip Sys Components). Licensee Program Meets Requirements of Item 2.1 (Part 1) of Generic Ltr 83-28 & Acceptable ML20245K3481989-08-11011 August 1989 Safety Evaluation Accepting Electrical Isolation Devices for Interfacing Safety & Nonsafety Sys Re Implementation of ATWS Rule ML20247H6421989-07-24024 July 1989 Safety Evaluation Granting 890118 Request for Relief from Hydrostatic Testing Requirements of Section XI of ASME Code ML20248C0851989-06-0202 June 1989 Safety Evaluation Supporting Amend 122 to License DPR-40 1999-09-02
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217B5401999-10-0606 October 1999 Safety Evaluation Supporting Amend 193 to License DPR-40 ML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data LIC-99-0096, Monthly Operating Rept for Sept 1999 for Fcs,Unit 1.With1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Fcs,Unit 1.With ML20211J9321999-09-0202 September 1999 Safety Evaluation Concluding That Licensee Proposed Alternatives Provide Acceptable Level of Quality & Safety. Proposed Alternatives Authorized for Remainder of Third ten- Yr ISI Interval for Fort Calhoun Station,Unit 1 LIC-99-0084, Monthly Operating Rept for Aug 1999 for Fort Calhoun Station.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Fort Calhoun Station.With ML20216E6431999-08-26026 August 1999 Rev 19 to TDB-VI, COLR for FCS Unit 1 ML20210R1961999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Fcs,Unit 1 ML20210G2181999-07-27027 July 1999 Safety Evaluation Supporting Amend 192 to License DPR-40 ML20210D9951999-07-22022 July 1999 Safety Evaluation Supporting Amend 191 to License DPR-40 ML20216E6361999-07-21021 July 1999 Rev 18 to TDB-VI, COLR for FCS Unit 1 ML20210R2081999-06-30030 June 1999 Revised Monthly Operating Rept for June 1999 for Fcs,Unit 1 LIC-99-0065, Monthly Operating Rept for June 1999 for Fort Calhoun Station,Unit 1.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Fort Calhoun Station,Unit 1.With ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20210P5461999-06-0808 June 1999 Rev 0,Vols 1-5 of Fort Calhoun Station 1999 Emergency Preparedness Exercise Manual, to Be Conducted on 990810. Pages 2-20 & 2-40 in Vol 2 & Page 4-1 in Vol 4 of Incoming Submittal Not Included ML20195B4581999-05-31031 May 1999 Rev 3 to CE NPSD-683, Development of RCS Pressure & Temp Limits Rept for Removal of P-T Limits & LTOP Requirements from Ts ML20207H7401999-05-31031 May 1999 Performance Indicators Rept for May 1999 LIC-99-0053, Monthly Operating Rept for May 1999 for Fort Calhoun Station,Unit 11999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Fort Calhoun Station,Unit 1 ML20195B4521999-05-17017 May 1999 Technical Data Book TDB-IX, RCS Pressure - Temp Limits Rept (Ptlr) ML20206L4241999-05-10010 May 1999 Safety Evaluation Supporting Corrective Actions to Ensure That Valves Are Capable of Performing Intended Safety Functions & OPPD Adequately Addressed Requested Actions Discussed in GL 95-07 ML20206M2601999-05-0606 May 1999 SER Concluding That Licensee IPEEE Complete Re Info Requested by Suppl 4 to GL 88-20 & IPEEE Results Reasonable Given FCS Design,Operation & History LIC-99-0047, Monthly Operating Rept for Apr 1999 for Fort Calhoun Station Unit 1.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Fort Calhoun Station Unit 1.With ML20195E8621999-04-30030 April 1999 Performance Indicators, for Apr 1999 ML20205Q5831999-04-15015 April 1999 Safety Evaluation Supporting Amend 190 to License DPR-40 ML20210J4331999-03-31031 March 1999 Changes,Tests, & Experiments Carried Out Without Prior Commission Approval for Period 981101-990331.With USAR Changes Other than Those Resulting from 10CFR50.59 ML20206G2641999-03-31031 March 1999 Performance Indicators Rept for Mar 1999 LIC-99-0034, Monthly Operating Rept for Mar 1999 for Fcs,Unit 1.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Fcs,Unit 1.With ML20205J8181999-02-28028 February 1999 Performance Indicators, for Feb 1999 LIC-99-0025, Monthly Operating Rept for Feb 1999 for Fort Calhoun Station,Unit 1.With1999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Fort Calhoun Station,Unit 1.With ML20207F3291999-01-31031 January 1999 FCS Performance Indicators for Jan 1999 ML20203B0991998-12-31031 December 1998 Performance Indicators for Dec 1998 LIC-99-0026, 1998 Omaha Public Power District Annual Rept. with1998-12-31031 December 1998 1998 Omaha Public Power District Annual Rept. with LIC-99-0003, Monthly Operating Rept for Dec 1998 for Fort Calhoun Station.With1998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Fort Calhoun Station.With ML20198S3771998-12-31031 December 1998 Safety Evaluation Supporting Amend 189 to License DPR-40 ML20198S4831998-12-31031 December 1998 Safety Evaluation Supporting Amend 188 to License DPR-40 ML20196G2251998-12-18018 December 1998 Rev 2 to EA-FC-90-082, Potential Over-Pressurization of Containment Penetration Piping Following Main Steam Line Break in Containment ML20198M3141998-11-30030 November 1998 Performance Indicators Rept for Nov 1998 LIC-98-0172, Monthly Operating Rept for Nov 1998 for Fort Calhoun Station,Unit 1.With1998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Fort Calhoun Station,Unit 1.With LIC-98-0160, Special Rept:On 981113,MSL RM RM-064 Was Declared Inoperable Due to Leakage Past Isolation Valve HCV-922.Troubleshooting Has Indicated That Leakage Has Stopped & Cause of Leak Continues to Be Investigated1998-11-25025 November 1998 Special Rept:On 981113,MSL RM RM-064 Was Declared Inoperable Due to Leakage Past Isolation Valve HCV-922.Troubleshooting Has Indicated That Leakage Has Stopped & Cause of Leak Continues to Be Investigated ML20203B0721998-11-16016 November 1998 Rev 6 to HI-92828, Licensing Rept for Spent Fuel Storage Capacity Expansion ML20196E4981998-10-31031 October 1998 Performance Indicators Rept for Oct 1998 ML20196G2441998-10-31031 October 1998 Changes,Tests & Experiments Carried Out Without Prior Commission Approval. with USAR Changes Other than Those Resulting from 10CFR50.59 LIC-98-0154, Monthly Operating Rept for Oct 1998 for Fort Calhoun Station,Unit 1.With1998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Fort Calhoun Station,Unit 1.With ML20154M4881998-10-19019 October 1998 Safety Evaluation Supporting Amend 186 to License DPR-40 ML20154N2411998-10-19019 October 1998 Safety Evaluation Supporting Amend 187 to License DPR-40 LIC-98-0136, Monthly Operating Rept for Sept 1998 for Fort Calhoun Station,Unit 1.With1998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Fort Calhoun Station,Unit 1.With ML20155G4261998-09-30030 September 1998 Performance Indicators for Sept 1998 ML20154A1251998-08-31031 August 1998 Performance Indicators, Rept for Aug 1998 LIC-98-0122, Monthly Operating Rept for Aug 1998 for Fort Calhoun Station Unit 1.With1998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Fort Calhoun Station Unit 1.With ML20238F7231998-08-17017 August 1998 Owner'S Rept for Isis ML18066A2771998-08-13013 August 1998 Part 21 Rept Re Deficiency in CE Current Screening Methodology for Determining Limiting Fuel Assembly for Detailed PWR thermal-hydraulic Sa.Evaluations Were Performed for Affected Plants to Determine Effect of Deficiency 1999-09-30
[Table view] |
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FORT CALHOUN STATION, UNIT 1 Docket No. 50-285 1.0 Introduction In the summer of 197'9, a pressurized water reactor (PWR) Licensee submitted a report to the NRC that identified a deficiency in its -
original analysis of_the containment pressurization resulting from a pos'tulated main steam line break (MSLB). A reanalysis of the containment pressure response folLowing a MSLB was performed, and it was determined that, if the auxiliary feedwater (AFW) system continued to supply feedwater at runout conditions to the steam generator that had experienced the steam line break, the containment design pressure would be exceeded in approximately 10 minutes. In l other words, the Long-term blowdown of the water supplied by the AFW system had not been considered in the earlier analysis.
On October 1, 1979, the foregoing information was provided to atL holders of operating. Licenses and construction permits in IE l
Information Notice 79-24 C23. Another Licensee performed an l accident analysis review pursuant to the information furnished in b
the above cited notice and discovered that, with offsite electrical power available, the condensate pumps would feed the affected steam
. generator at an excessive rate. This excessive feed had not been considered in its analysis of the postulated MSLB accident.
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A third Licensee informed the NRC of an error in the MSLB analysis for their plant. For a zero or low power condition at the en'd of -
core Life, the licensee identified an incorrect po'stulation that the startup feedwater control valves wouLd remain positioned "as is" during the transient. In reality, the startup feedwater control valves wilL ramp to 80% full open due to an override signal resulting from the Low steam generator pressure reactor trip signal.
Reanalysis of the events showed that the rate of feedwater addition to the affected steam generator associated with the opening of the startup valve would cause a rapid reactor cooldown and resultant reactor-return-to power response, a condi, tion which is beyond the plant's design basis.
FolLowing the identification of these deficiencies in the original l MSLB accident analysis, the NRC issued IE Bulletin 80-04 on l
February 8, 1980. This butLetin required alL Licensees of PWRs and certain near-term PWR operating License applicants to do the folLowing:
"1. Review the containment pressure response analysis to determine if the potential for containment overpressure for MSLB inside
. containment included the impact of runout flow from the auxiliary feedwater system and the impact of other energy sources such as continuation of feedwater or condensate flow.
In your review, consider your ability to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at runout flow.
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"2. Review your analysis of the reactivity increase which results from a MSLB inside or outside containment. This review *should -
consider the reactor cooldown rate and the potential for the reactor to. return to power with the most reactive control rod in the fully withdrawn position. If your previous analysis did not consider atL potential water sources (such as those Listed in 1 above) and.if the reactivity increase is greater than previous analysis indicated, the report of this review should include: ,
.a. The boundary conditions for the analysis, e.g., the end of life shutdown margin, the moderqtor temperature coefficient, power level and the net effect of the associated steam generator water inventory on the reactor system cooling, etc.;
- b. The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution.to the reactor coolant system;
- c. The effect of extended water supply to the affected steam generator on the core criticality and return to power; and
( d. The hot channel factors corresponding to the most reactive rod in the fully withdrawn positions at the end of Life, and the Minimum Departure from Nucleate Boiling Ratio (MDNBR) values for the analyzed transient.
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- 3. If the potential for containment overpressure exists or the reactor return-to power response worsens, provide a proposed -
corrective action and a schedule for completion of the corrective. action. If the unit is operating, provide a description of any interim action that will be taken until the proposed corrective action i s completed."
Following the licensee's initial response to IE Bulletin 80-04, a .
request for additional information was developed to obtain all the information necessary to evaluate the licensee's analysis.
The results of our evaluation for Fort Calhoun Station, Unit 1 (Fort Calhoun 1) are provided below.
2.0 Evaluation _
Our consultant, the Franklin Research Center (FRC), has. reviewed the submittals made by the Omaha Public Power District in response to IE Bulletin 80-04, and prepared the attached Technical Evaluation Report. We have reviewed this evaluation and concur in l
its bases and findings.
i l
l 3.0 Conclusion Based on our review of the enclosed Technical Evaluation Report, the following conclusions are made regarding the postulated MSLB with continued feedwater addition for Fort Calhoun 1:
, 1. There is no potential for containment overpressurization resulting from a MSLB with continued feedwater addition
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1 because the main feedwater system is isolated and the auxiliary feedwater actuation system prevents actuation.of the ,,
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affected steam generator from being fed; '
- 2. The AFW pumps will not experience runout conditions; therefore, they will be able to carry out their intended function without incurring damage during a MSLB;
- 3. ALL potential water sources were identified and, although a reactor return-to power due to decay heat and subcritical multiplication occurs, the reactor remains suberitical, and the DNOR remains greater than 1.30; therefore, the Reference 3 reactivity increase analysis remains valid;
- 4. No further action is required of the licensee regarding IE Bulletin 80-04; and
- 5. Compliance of the Main Steam Isolation Signal, Containment Isolation Actuation Signal and the Safety Injection Actuation Signal with IEEE Standard 279-1971 was not considered in this review.
I .
l 4.0 References
- 1. " Analysis of a PWR Main Steam Line Break with continued Feedwater Addition," NRC Office of Inspection and Enforcement, Februa ry 8,1980, IE Bulletin 80-04
- 2. "Overpressurization of the Containment of a PWR Plant after 1
a Main Steam Line Break," NRC Office of Inspection and l
Enforcement, O c tobe r 1,1979, IE Information Notice 79-24
- 3. W. C. Jones (OPPD) letter to K. V. Seyfrit (NRC, Region IV),
Subject:
IE Bulletin 80-04, Response to Item 2, May 15,1980 1- .
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4
- 4. W. C. Jones (0 PPD), letter to K. V. Seyfrit (NRC, Region IV),
Subject:
IE Bulletin 80-04, Response to Item, August 27, 1980 .
Jones (OPPD) letter to H. R. Denton (NR'C) ,
Subject:
- 5. W. C.
Application for Amendment of Operating License, November 17, 1981
- 6. Fort Calhoun Station, Unit 1 Final Safety Analysis Report, through Rev. 27, Omaha Public Power District .
- 7. "PWR Main Steam Line Break with Continued Feedwater Addition - ,
Review of Acceptance Criteria," Franklin Research Center, November 17,1981, TER-C5 506-119
- 8. " Criteria for Protection Systems for, Nuclear Power Generating Stations," Institute of Electrical and Electronics Engineers, New York, NY, 1971, IEEE Std 279-1971
- 9. Standard Review Plan, S e ct i on 15.1.5, " Steam System Piping Failures Inside and Outside of Containment (PWR)", NRC, July 1981, NUREG-0800
- 10. " Criteria for Accident Monitoring Functions in Light-Water Cooled Reactors," American Nuclear Society, Hinsdale, IL, December 1980, ANS/ ANSI-4.5-1980
- 11. " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Rev., NRC, D e c embe r 1980, Regulatory Guide 1.97
- 12. " Single Failure Criteria for PWR Fluid Systems g" American Nuclear Society, Hinsdale, IL, June 1976, ANS-51.7/658-1976
- 13. " Quality G roup Classification and Standards for Water , Steam ,
3 and Radioactive-Waste-Containing Components of Nuclear Power Plants," Rev. 3, NRC, February 1976, Regulatory Guide 1.26 a
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- 14. " Interim Staff Position on Environmental Qualification of
. Safety-Related Electrical Equipment," Rev. 1, N R C , J u L y .19 81, ,,
NUREG-0588
- 15. " Auxiliary Feedwater System Automatic Initiation and Flow Indication, Fort Calhoun Station," D e c embe r 8, 1981, TER-C5257-297
- 16. R. A. CL6rk (NRC) letter to N. C. Jones (OPPD),
Subject:
Auxiliary Feedwater System, February 20, 1981.
Attachment:
Franklin Research Center Technical Evaluation' Report 9
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