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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20210B7941999-07-15015 July 1999 SER Denying Licensee Request for Changes to Current Ts,Re Deletion of Tendon Surveillance Requirements for Calvert Cliffs ML20196C6981999-06-21021 June 1999 Safety Evaluation Concluding That Use of ASME Section XI Code Including Summer 1983 Addenda as Interim Code for Third 10-year Insp Interval at Calvert Cliffs Units 1 & 2 Until Review of 1998 Code Completed,Would Be Acceptable ML20204H6471999-03-21021 March 1999 SER Re License Renewal of Calvert Cliffs Nuclear Power Plant,Units 1 & 2 ML20203D4311999-02-0505 February 1999 Safety Evaluation Accepting Procedure Established for long-term Corrective Action Plan Related to Containment Vertical Tendons ML20199G4671999-01-20020 January 1999 SER Accepting USI A-46 Implementation for Plant ML20199F4781999-01-0808 January 1999 Safety Evaluation Concluding That Bg&E Performed Appropriate Evaluations of Operational Configurations of safety-related power-operated Gate Valves to Identify Valves Susceptible to Pressure Locking.Concludes GL 95-07 Actions Were Addressed ML20198S7591999-01-0707 January 1999 SER Accepting Quality Assurance Program Description Change for Calvert Cliffs Nuclear Power Plant,Units 1 & 2 ML20195H1001998-11-16016 November 1998 Safety Evaluation of First Containment Insp Interval Iwe/Iwl Program Alternative ML20154G3931998-10-0505 October 1998 Safety Evaluation Concluding That Flaw Tolerance Evaluation for Assumed Flaw in Inboard Instrument Weld of Pressurizer Meets Rules of ASME Code ML20237D4981998-08-19019 August 1998 Safety Evaluation Accepting Licensee Request for Extension of Second ten-year Inservice Insp Interval ML20236L7521998-07-0606 July 1998 Safety Evaluation Granting Bg&E 980527 Request for Relief from Requirement of Section IWA-5250 of ASME Code for Calvert Cliffs Unit 2.Alternatives Provide Reasonable Assurance of Operational Readiness ML20236F7791998-06-30030 June 1998 Safety Evaluation Authorizing Request for Temporary Relief from Requirement of Subsection IWA-5250 of ASME Code,Section XI for Plant,Unit 1 ML20217Q7971998-05-0404 May 1998 Safety Evaluation Supporting Amends 227 & 201 to Licenses DPR-53 & DPR-69,respectively ML20217P7441998-04-0707 April 1998 SE Accepting Licensee Request for Second 10-yr Insp Interval to Use ASME Section IX Code Case N-416-1, Alternative Pressure Test Requirement for Replacement Items by Welding, Class 1,2 & 3,Div 1 at Plant Per 10CFR50.55a(a)(3)(i) ML20203E9741998-02-11011 February 1998 SER Accepting IST Program Requests for Relief Submitted for Third 10-yr Intervals for Pumps & Valves for Plant ML20198K3741998-01-0505 January 1998 Safety Evaluation Authorizing Second 10-yr ISI Interval for Plant,Units 1 & 2 Relief Request from ASME Boiler & Pressure Vessel Code Requirements to Use Performance Demonstration Initiative Program ML20197G1291997-12-15015 December 1997 Safety Evaluation Accepting Revised Withdrawal Schedules for Plant,Units 1 & 2 ML20134G5291997-02-0707 February 1997 Safety Evaluation Re Request Relief from ASME Code Requirement for Calvert Cliffs Nuclear Power Plant,Unit 1 ML20058G0561993-11-29029 November 1993 Safety Evaluation Granting Relief from Installation Requirements of Section Iii,Article 9 of 1968 Edition of ASME Boiler & Pressure Vessel Code for Plant ML20057C3831993-09-22022 September 1993 Sser Accepting Licensee 930707 Response to Station Blackout Rule,Per 10CFR50.63 ML20127C0571992-12-23023 December 1992 Safety Evaluation Accepting Clarification of Previous Reactor Vessel Surveillance Program Actions & Withdrawal Schedule Change for Reactor Matl Surveillance Capsule ML20128B9371992-11-20020 November 1992 Safety Evaluation Accepting Licensee Response to Suppl 1 to GL 87-02 IR 05000317/19900801990-11-21021 November 1990 SER Determining That Load Profile Used During Surveillance Test Procedure Meets Requirement of Tech Spec 4.8.2.3.2.e, Per Unresolved Item Noted in Insp Repts 50-317/90-80 & 50-318/90-80 ML20059N2251990-10-0101 October 1990 Safety Evaluation Accepting Proposed Actions Re Potential Design Deficiency,Per LER 89-023 ML20245E8561989-06-22022 June 1989 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28, Reactor Trip Sys Reliability for All Domestic Operating Reactors ML20247D0361989-03-24024 March 1989 Safety Evaluation Supporting Amend 119 to License DPR-69 ML20205P6491988-11-0202 November 1988 Safety Evaluation Re Compliance w/10CFR50.62 ATWS Rule. Auxiliary Feedwater Actuation Sys Designs in Conformance W/ Requirements of TMI Action Plan Item II.E.1.2,surveillance Testing,Complies W/Atws Rule ML20238A3141987-08-24024 August 1987 Safety Evaluation of Encl Technical Evaluation Repts EGG-NTA-7243,EGG-NTA-7720 & EGG-NTA-7463 & Licensee Submittals in Response to Generic Ltr 83-28,Items 2.1.1, 2.1.2 & 4.5.2.Responses Acceptable ML20213E8731987-05-11011 May 1987 Safety Evaluation Supporting Util 861002 Request for Relief from Inservice Insp Requirements of ASME Code Section XI Re Peripheral CRD Housings ML20205R7061987-03-26026 March 1987 Safety Evaluation Supporting Relief from Certain ASME Code Hoop Stress Requirements Until Next Refueling Outage to Repair Flaw in Base Metal of Steam Generator Main Steam Line 12 ML20210P0911987-02-0404 February 1987 Safety Evaluation Re Projected Values of Matl Properties for Fracture Toughness Requirements for Protection Against PTS Events.Util Must Submit Reevaluation of Resistance Temp PTS ML20207P0721987-01-0606 January 1987 Safety Evaluation Supporting Conformance to Reg Guide 1.97 W/Noted Exceptions ML20198G6971986-05-22022 May 1986 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 1.2 Re post-trip Review Data & Info Capability ML20210P2801986-05-0606 May 1986 Safety Evaluation Accepting Licensee Response to Items 3.1.3 & 3.2.3, Post-Maint Testing of...Safety-Related Components, of Generic Ltr 83-28 ML20202E0721986-03-31031 March 1986 Safety Evaluation Supporting Amend 116 to License DPR-53 ML20141N7691986-03-10010 March 1986 Safety Evaluation Granting Util 850913 & 860113 Requests for Relief from ASME Boiler & Pressure Vessel Code,Section XI Re Observation of Lubricant Level or Pressure During Pump Testing ML20136G7371985-11-14014 November 1985 Safety Evaluation Granting Relief from ASME Boiler & Pressure Vessel Code,Section XI Requirement Re Pressure Testing of Class 2 Piping Associated W/Rcs ML20138E2821985-10-10010 October 1985 Safety Evaluation Re Extended Burnup of C-E PWR Fuel. Methodologies Acceptable for Extended Burnup Application ML20137W8811985-09-18018 September 1985 Safety Evaluation Granting Relief from Radiographic & Visual Insp of Reactor Coolant Pump Casings ML20126L8241985-07-18018 July 1985 Suppl to Safety Evaluation Supporting Response to IE Bulletin 80-11 Re Reevaluation of 22 Masonry Walls,Utilizing Methods Other than Energy Balance Technique ML20133E1101985-07-18018 July 1985 Corrected Suppl to Safety Evaluation Accepting Util 850612 Reevaluation of Masonry Wall Design Using Linear Elastic Methodology.Action Completed on IE Bulletin 80-11 ML20205B2591985-07-10010 July 1985 Safety Evaluation Accepting Licensee Position on Generic Ltr 83-28 Items 4.2.1 & 4.2.2 Re Reactor Trip Sys Reliability ML20127N2351985-05-0909 May 1985 SER Re NUREG/CR-3511, Interim Reliability Evaluation Program:Analysis of Calvert Cliffs Unit 1 Nuclear Power Plant. Use of PRA Results Out of PRA Context Clearly Inappropriate ML20132B8731984-03-21021 March 1984 SER on Multi-Plant Action C-14 Re Seismic Qualification of Auxiliary Feedwater Sys ML20148A9641978-01-0404 January 1978 Safety Evaluation Supporting Amends 27 & 12 to Licenses DPR-53 & DPR-69,respectively 1999-07-15
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G6971999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Calvert Cliffs Npp,Units 1 & 2.With ML20216J8731999-09-10010 September 1999 Rev 52 to QA Policy for Calvert Cliffs Nuclear Power Plant ML20212A4441999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Ccnpp,Units 1 & 2. with ML17326A2011999-08-23023 August 1999 LER 99-004-00:on 990724,reactor Tripped Due to Main Transformer Bushing Flashover.Plant Was Brought to SS & Components Were Tested & Performed Satisfactorily.With 990823 Ltr ML20210S6091999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Ccnpp,Units 1 & 2. with ML20210N6001999-07-27027 July 1999 ISI Summary Rept for Calvert Cliffs Unit 2. Page 2 of 3 in Encl 1 of Incoming Submittal Not Included ML20210B7941999-07-15015 July 1999 SER Denying Licensee Request for Changes to Current Ts,Re Deletion of Tendon Surveillance Requirements for Calvert Cliffs LD-99-039, Part 21 Rept Re Defect of Abb 1200A 4kV Vacuum Breakers. Initially Reported on 990625.Defect Results in Breaker Failing to Remain in Closed Position.Root Cause Evaluation & Corrective Action Plan Being Developed.Licensee Notified1999-06-30030 June 1999 Part 21 Rept Re Defect of Abb 1200A 4kV Vacuum Breakers. Initially Reported on 990625.Defect Results in Breaker Failing to Remain in Closed Position.Root Cause Evaluation & Corrective Action Plan Being Developed.Licensee Notified ML20209F1721999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Calvert Cliffs Npp.With LD-99-035, Part 21 Rept Re Abb 1200A 4KV Vacuum Breakers Performing Trip Free Operation When Close Signal Received by Breaker. Defect Results in Breaker Failing to Remain in Closed Position.Root Cause & CAP Being Developed1999-06-25025 June 1999 Part 21 Rept Re Abb 1200A 4KV Vacuum Breakers Performing Trip Free Operation When Close Signal Received by Breaker. Defect Results in Breaker Failing to Remain in Closed Position.Root Cause & CAP Being Developed ML20196C6981999-06-21021 June 1999 Safety Evaluation Concluding That Use of ASME Section XI Code Including Summer 1983 Addenda as Interim Code for Third 10-year Insp Interval at Calvert Cliffs Units 1 & 2 Until Review of 1998 Code Completed,Would Be Acceptable ML20195K2811999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Ccnpp,Units 1 & 2. with ML20206R5871999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Ccnpp,Units 1 & 2. with ML20195B3891999-04-30030 April 1999 0 to CENPD-279, Annual Rept on Abb CE ECCS Performance Evaluation Models ML20205N2951999-04-13013 April 1999 Special Rept:On 990314,fire Detection Sys Was Removed from Svc to Support Mod to Replace SRW Heat Exchangers in Unit 2 SRW Room During Unit 2 Refueling Outage.Contingency Measure 15.3.5.A.1 Will Continue Until Fire Detection Sys Restored ML20210T5211999-04-0101 April 1999 Rev 0 to Ccnpp COLR for Unit 2,Cycle 13 ML20205P5441999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Calvert Cliffs Nuclear Power Plant,Units 1 & 2.With ML20204H6471999-03-21021 March 1999 SER Re License Renewal of Calvert Cliffs Nuclear Power Plant,Units 1 & 2 ML20207M8321999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Calvert Cliffs Nuclear Power Plant.With ML20203D4311999-02-0505 February 1999 Safety Evaluation Accepting Procedure Established for long-term Corrective Action Plan Related to Containment Vertical Tendons ML20199G4671999-01-20020 January 1999 SER Accepting USI A-46 Implementation for Plant ML20206Q3221999-01-11011 January 1999 Special Rept:On 981226,wide Range Noble Gas Effluent RM Was Removed from Operable Status.Caused by Failure of mid-range Checksource to Properly Reseat.Completed Maint & post-maint Testing & RM Was Returned to Operable Status on 990104 ML20207L0451999-01-0808 January 1999 Cost-Benefit Risk Analyses:Radwaste Sys for Light Water Reactors ML20199F4781999-01-0808 January 1999 Safety Evaluation Concluding That Bg&E Performed Appropriate Evaluations of Operational Configurations of safety-related power-operated Gate Valves to Identify Valves Susceptible to Pressure Locking.Concludes GL 95-07 Actions Were Addressed ML20198S7591999-01-0707 January 1999 SER Accepting Quality Assurance Program Description Change for Calvert Cliffs Nuclear Power Plant,Units 1 & 2 ML20207M2281998-12-31031 December 1998 1998 Annual Rept for Bg&E. with ML20199E2931998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Calvert Cliffs Npp. with ML20206R9911998-12-0808 December 1998 Rept of Changes,Tests & Experiments (10CFR50.59(b)(2)). with ML20198B2631998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Calvert Cliffs Nuclear Power Plant,Units 1 & 2.With ML20195H1001998-11-16016 November 1998 Safety Evaluation of First Containment Insp Interval Iwe/Iwl Program Alternative ML20196E2211998-10-31031 October 1998 Non-proprietary Rev 03-NP to CEN-633-NP, SG Tube Repair for Combustion Engineering Designed Plant with 3/4 - .048 Wall Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves ML20195E5281998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Calvert Cliffs Nuclear Power Station,Units 1 & 2.With ML20154Q7191998-10-21021 October 1998 Special Rept:On 980923,unit 1 Wrngm Was Removed from Operable Status.Caused by Failure of Process Flow Transducer.Completed Maint to Remove Process Flow Transducer Input to Wrngm Microprocessor & Completed Formal Evaluation ML20154G3931998-10-0505 October 1998 Safety Evaluation Concluding That Flaw Tolerance Evaluation for Assumed Flaw in Inboard Instrument Weld of Pressurizer Meets Rules of ASME Code ML20154M5841998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Calvert Cliffs Nuclear Plant,Units 1 & 2.With ML20153C2571998-09-18018 September 1998 Special Rept:On 980830,wide Range Noble Gas Monitor (Wrngm) Channel Was Removed from Operable Status.Caused by Need to Support Performance of Required 18-month Channel Calibr.Will Return Wrngm to Operable Status by 980925 ML20153C1091998-09-18018 September 1998 Part 21 Rept Re Defective Capacity Control Valves.Trentec Personnel Have Been in Contact with Bg&E Personnel Re Condition & Have Requested Potentially Defective Valves ML20151U5441998-09-0404 September 1998 Bg&E ISI Summary Rept for Calvert Cliffs ML20151T5281998-09-0101 September 1998 Special Rept:On 980819,declared Rv Water Level Monitor Channel a Inoperable.Caused by Failure of Three Heated Junction Thermocouples (Sensors) in Lower Five Sensors. Channel a & B Rv Water Level Probes Will Be Replaced ML20151Y1191998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Calvert Cliffs Nuclear Power Plant Units 1 & 2.With ML20237D4981998-08-19019 August 1998 Safety Evaluation Accepting Licensee Request for Extension of Second ten-year Inservice Insp Interval 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 ML20237B9371998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Calvert Cliffs Nuclear Power Plant ML20237D5941998-07-22022 July 1998 Rev 2 to Ccnpp COLR for Unit 2,Cycle 12 ML20236L7521998-07-0606 July 1998 Safety Evaluation Granting Bg&E 980527 Request for Relief from Requirement of Section IWA-5250 of ASME Code for Calvert Cliffs Unit 2.Alternatives Provide Reasonable Assurance of Operational Readiness ML20236F7791998-06-30030 June 1998 Safety Evaluation Authorizing Request for Temporary Relief from Requirement of Subsection IWA-5250 of ASME Code,Section XI for Plant,Unit 1 ML20236R0881998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Calvert Cliffs Nuclear Power Plant,Units 1 & 2 ML20236X3101998-06-19019 June 1998 Rev 1 to Calvert Cliffs Nuclear Power Plant COLR for Unit 2,Cycle 12 ML20249A9571998-06-15015 June 1998 Special Rept:On 980430,fire Detection Sys Was Removed from Svc to Support Mod to Purge Air Sys 27-foot Elevation & 5-foot Elevation East Piping Penetration Rooms.Installed Temporary Alteration & Returned Fire Detection Sys to Svc ML20249A7711998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Ccnpp,Units 1 & 2 1999-09-30
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5( UNITED STATES
[ p, NUCLEAR REGULATORY COMMISSION L -l WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION INILRIM RELIABILITY EVALUATION PROGRAM (IREP) REPORT BALTIMORE GA5 AND ELECTRIC COMPANY CALVERI CLIFF 5 UNIT NO. 1 DOCKET NO. 50-317 Introduction The report entitled " Interim Reliability Evaluation Program: Analysis of the Calvert Cliffs Unit 1 Nuclear Power Plant," NUREG/CR-3511, March 1984, is an evaluation of Calvert Cliffs Unit I using probabilistic risk assessment (PRA) techiques. The IREP was undertaken by Sandia National Laboratory for the Nur str Regulatory Commission in order to further the application of PRA tr ques to nuclear reactor safety analyses. Although the IREP was not i.... ted as a result of any particular safety concern, the results of the study did yield several interesting insights into the Calvert Cliffs facility, including the identification of significant single failures.
Discussion and Evaluation The Calvert Cliffs IREP PRA analysis used fault tree and event tree models as the primary tools to evaluate the risk due to a core melt at Calvert Cliffs Unit 1. Core melt sequences initiated by one of three break-size loss-of-
! coolant accidents (LOCAs) or one of six categories of transients were evaluated, and the dominant (i.e., highest frequency) sequences were further analyzed to estimate the magnitude and frequency of radionuclide release.
The most significant sequences contributing to the core melt frequency are (1)Antici l frequency) pated Transients Without Scram (ATWS) (33% of the total core melt, (2) l system failure in the recirculation phase (20% of the total core melt frequency), (3) the loss of a DC bus followed by failure of secondary heat removal (16% of the total core melt frequency), and (4) Loss of Offsite Power followed by failure of secondary heat removal or a stuck open relief valve l (12% of the total core melt frequency). Transients involvin l
power conversions and secondary steam release systems (6% (PCS) g (1)core of the total loss of melt frequency), (2) transients requiring primary system pressure relief not including ATWS (1% of the total core melt frequency) and (3) all other l transients (5% of the core melt frequency), were judged to be of lesser i
significance.
l By letter dated January 22, 1985, the staff requested that BG&E review the IREP report and provide comments. The response by BG&E, in their letter dated l
March 26, 1985, indicated that they had reviewed NUREG/CR-3511 and had made certain system / procedural changes.
- In the sequence involving loss of PCS followed by loss of auxiliary feedwater (AFW),the mispositioning of valve AFW-161 was the most significant contributor. '
l This valve is located at the condensate storage tank discharge.
1 OO 7 R
P
~
The licensee indicated that this valve at both units (1-AFW-161 and-2-AFW-161) had been modified by removing the valve internals, thus eliminating this potential failure. In addition, the licensee indicated that procedures to use the AFW cross connect valve (MOV-4550) had been incorporated into Abnormal Operating Procedure 3D,Section III, thus decreasing the likelihood of total loss of AFW.
For large-break LOCAs, recirculation of coolant from the containment sum the reactor coolant system via the high pressure safety injection (HPSI)p to pumps is credited. In the event of HPSI failure, the low pressure safety injection (LPSI) pumps may be used if reactor coolant system pressure i is sufficiently lowered. Use of LPSI would require reset of the
- . recirculation actuation signal (RAS) since RAS trips the LPSI pumps. The i licensee has installed a key locked switch to override the RAS signal to the selected LPSI pump to facilitate post-RAS use of LPSI pumps.
For LOCAs, the salt water system (SWS) provides the ultimate heat sink
- (ChesapeakeBay.) Equipment which must function post-LOCA and which
- requires cooling is serviced by one of two systems (Service Water System
- - (SRWS) or Component Cooling Water System (CCWS) which in turn rejects heat to the SWS via two CCWS or.two SRWS heat exchangers.) The ECCS Room coolers reject heat directly to the SWS. The common discharge on the salt water side of both-CCWS heat exchangers passes through valve SWS-196; failure of SWS-196 in the closed position would result in loss of containment spray, HPSI, and LPSI. Similarly, the common discharge on the salt water side of the SRWS heat exchanger passes through SWS-197; failure of SWS-197 in the closed position would result in loss of diesel generators (during a LOCA with loss
- of offsite power) and the containment fan coolers. Accordingly, valves
- SWS-196 and SWS-197 at Calvert Cliffs Units 1 and 2 have been modified by removal of their internals to prevent failure in the closed position.
c
- For the station blackout event, it is assumed that both off-site and onsite
. power are unavailable. Core melt would result from failure of secondary
! heat removal. Calvert Cliffs has an available NRC-approved, alternate source
! of off-site power which consists of a 69KV line from the Southern Maryland Electrical Cooperative (SMECO). The IREP report states that,'"In light of
, the importance of loss of offsite power sequences, in general, and station l blackout in particular, the utility is reviewing its procedure for connecting the 69KV line. Also, as a result of Task Action Plan A-44 Station Blackout, it is likely that all plants will be required to have improved loss of offsite power procedures. These improved procedures and other changes should result in significant mitigation of loss of offsite power sequences in the future."
Procedures for use of the 69KV SMECO line have been developed by BG&E and incorporated in Operating Instruction 27E,Section IV.
Although some of the system / procedural changes described in BG8E's March 26,
, 1985 letter were developed as a result of other programs, these changes are consistent with the IREP findings and result in improvements to the safety of the Calvert Cliffs facility.
i 1
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Y Conclusion There was nothing found in the Calvert Cliffs PRA results to indicate a need for NRC action. There are no apparent " outlier" risk dominant sequences in the PRA that suggest such a need. The core melt frequency for Calvert Cliffs Unit I was, in fact, similar to values predicted by PRA for other pressurized water reactors. Improvements made to Calvert Cliffs, including those described herein, decrease the core melt frequency still further.
With regard to the usefulness of the IREP results, any assessment of facility safety should be made in light of the overall limitations of the PRA technique; thus, use of PRA results out of PRA context is clearly inappropriate.
I Principal Contributor:
D. Jaffe Dated: May 9, 1985 2
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