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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML20199L5891997-12-24024 December 1997 Independent Analysis & Evaluation of AM-241 & Transuranics & Subsequent Dose to Two Male Workers at Connecticut Yankee Atomic Power Plant ML20210P8721997-08-31031 August 1997 Post Decommissioning Activities Rept, for Aug 1997 ML20141A0041997-05-31031 May 1997 Independent Assessment of Radiological Controls Program at Cyap Haddam Neck Plant Final Rept May 1997 ML20137H3031997-03-31031 March 1997 Rev 2 to Nuclear Training Loit/Lout Audit Reviews ML20137C6281997-03-14014 March 1997 Redacted Version of Rev 1 to Nuclear Training Loit/Lout Audit Reviews ML20135E3221997-02-13013 February 1997 Independent Review Team Rept 1996 MP -1 Lout NRC Exam Failures ML20134L2751997-02-0303 February 1997 Draft Rev to GPRI-30, Spent Fuel Storage Facility Licensing Basis/Design Basis ML20134L2911997-01-0808 January 1997 Rev 0 to UFSAR Rev Grpi ML20237E0751996-12-31031 December 1996 Decommissioning Cost Study for Connecticut Yankee Nuclear Power Plant ML20204B6811996-12-31031 December 1996 Nuclear Lemons Assessment of America Worst Commerical Nuclear Power Plants. 5th Edition ML20134L2591996-12-31031 December 1996 Rev 0 to Sys Reclassification ML20134L2721996-12-31031 December 1996 Commitment Mgt Grpi ML20134L2481996-12-30030 December 1996 Rev 0 to Sys Needed for Decommissioning Sys Evaluation Process ML20135E2711996-12-26026 December 1996 Evaluation of Intakes by Two Workers at CT Yankee Atomic Power Co ML20134L2841996-12-23023 December 1996 Rev 0 to Spent Fuel Pool Island Studies ML20134L2381996-12-19019 December 1996 Rev 0 to Accident Analysis ML20134L2871996-12-19019 December 1996 Rev 0 to Grpi for Defueled Condition Fire Protection Technical Requirements ML20058G3481993-10-31031 October 1993 Nonproprietary VIPRE/WRB-1 DNBR Thermal Limit for Westinghouse Fuel Types, for Oct 1993 ML20058G6771993-09-29029 September 1993 Rev 0 to Project Assignment 91-006, Connecticut Yankee Event V Mods ML20058L7411993-07-15015 July 1993 QA Svcs Dept Surveillance Rept SS-169, Assessment of Fitness for Duty Program ML20127J5391992-11-30030 November 1992 Connecticut Yankee Structural Reevaluation of Plant Structures to Address SEP Topic III-7.B Load Combinations Code Changes ML20125E5661992-03-31031 March 1992 Nonproprietary PWR SG Tube Repair Limits:Technical Support Document for Expansion Zone PWSCC in Roll Transitions Final Rept A08995, Rev 1 to Auxiliary Initiation Event Analysis1990-09-14014 September 1990 Rev 1 to Auxiliary Initiation Event Analysis ML20062G4951990-07-31031 July 1990 Rev 3 to Northeast Utils USNRC Read & Sign ML20055G4951990-07-31031 July 1990 Decommissioning Financial Assurance Certification Rept ML20055D8931990-06-30030 June 1990 Fracture Mechanics Evaluation:Haddam Neck Pressurizer ML20246N9421989-08-31031 August 1989 Tornado Missile Risk Analysis of Bleed & Feed & Auxiliary Feedwater Safe Shutdown Sys at Connecticut Yankee Atomic Power Station ML20206J9991988-10-31031 October 1988 Crdr Summary Rept, Per Suppl 1 to NUREG-0737 ML20153G4081988-06-0303 June 1988 Rev 0 to Conneticut Yankee Replace Nuclear Instrumentation Sys, Neo Project Description ML20153G4031988-03-0808 March 1988 Rev 0 to, Connecticut Yankee Modernize Reactor Protection Sys - Phase 2, Conceptual Project Description ML20235B5911987-09-24024 September 1987 Undated Technical Evaluation:Haddam Neck Plant Reanalysis of Non-LOCA DBAs & Tech Spec Change Requests in Support of Reload Cycle 15 B12619, Bimonthly Progress Rept 5:New Switchgear Bldg Const1987-07-31031 July 1987 Bimonthly Progress Rept 5:New Switchgear Bldg Const ML20214R6371987-06-30030 June 1987 Technical Rept Supporting Cycle 15 Operation ML20205Q7011987-04-30030 April 1987 Bimonthly Progress Rept 3:New Switchgear Bldg Const B12480, Const Design Info Submittal:New Switchgear Bldg Const. W/38 Oversize Encl1987-04-30030 April 1987 Const Design Info Submittal:New Switchgear Bldg Const. W/38 Oversize Encl ML20209C7161986-12-31031 December 1986 Vols 1 & 2 to Isap,Haddam Neck Plant,Final Rept ML20238C3821986-12-31031 December 1986 Connecticut Yankee 1986 Steam Generator Repair Outage ALARA Rept ML20238C3751986-12-31031 December 1986 Connecticut Yankee 1986 Core XIII - Xiv Refueling Outage ALARA Rept ML20202D5071986-06-30030 June 1986 Non-LOCA Transient Analysis ML20206J8871986-06-30030 June 1986 Isap Public Safety Impact Model Project Analyses Summaries ML20205T1201986-05-19019 May 1986 Steam Generator Tube Insp Rept for 1986 Refueling Outage ML20205T1231986-05-0808 May 1986 Steam Generator Tube W/55% Through Wall Defect Safety Evaluation for Structural Integrity CY-86-031, Rev 1 to 4.10.1(D) Inservice Insp of Steam Generator Tubes (Acceptance Criteria)1986-05-0808 May 1986 Rev 1 to 4.10.1(D) Inservice Insp of Steam Generator Tubes (Acceptance Criteria) ML20138B2001986-03-31031 March 1986 Status of Containment Integrated Leak Rate Test Program ML20138B1831986-03-31031 March 1986 Schedular Exemption Requests Re Type C Leak Testing ML20138B1561986-03-31031 March 1986 Permanent Exemptions Re Types a & C Leak Testing ML20199L0731986-03-31031 March 1986 Probabilistic Safety Study,Northeast Util Svc Company Evaluation of Probabilistic Safety Study Results B12020, Summary of Results of Haddam Neck Probabilistic Safety Study1986-03-31031 March 1986 Summary of Results of Haddam Neck Probabilistic Safety Study ML20155E7591986-02-28028 February 1986 Best Estimate LOCA Analysis, Vol 5 CY-85-062, Rev 2 to Steam Generator Tube Sleeving & Plugging1986-01-22022 January 1986 Rev 2 to Steam Generator Tube Sleeving & Plugging 1997-08-31
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217K3301999-10-19019 October 1999 Safety Evaluation Supporting Amend 195 to License DPR-61 ML20206C8761999-04-28028 April 1999 Safety Evaluation Supporting Amend 194 to License DPR-61 CY-99-047, Ro:On 981217,identified Unsuccessful Dewatering of Cnsi HIC, Model PL8-120R,containing Resins.Caused by Apparent Failure of Dewatering Tree.Other HICs Have Been Procured,Recertified & Returned to Plant for Use1999-03-23023 March 1999 Ro:On 981217,identified Unsuccessful Dewatering of Cnsi HIC, Model PL8-120R,containing Resins.Caused by Apparent Failure of Dewatering Tree.Other HICs Have Been Procured,Recertified & Returned to Plant for Use ML20206F1971998-12-31031 December 1998 Annual Rept for 1998 for Cyap. with CY-99-027, Annual Rept for 10CFR50.59, for Jan-Dec 1998.With1998-12-31031 December 1998 Annual Rept for 10CFR50.59, for Jan-Dec 1998.With ML20198G9101998-12-22022 December 1998 Proposed Rev 2 of Cyap QAP for Haddam Neck Plant. Marked Up Rev 1 Included ML20238F2131998-08-28028 August 1998 SER Accepting Defueled Emergency Plan for Emergency Planning for Connecticut Yankee Atomic Power Co CY-98-136, Ro:On 980727,flow Blockage Occurred & Caused Pressure in Sys to Increase,Resulting in Relief Valve Lifting & Pipe Vibration,Which Caused Leaks to Develop.Caused by Nearly Closed post-filter Inlet Valve.Repaired 2 Leaks in Line1998-08-12012 August 1998 Ro:On 980727,flow Blockage Occurred & Caused Pressure in Sys to Increase,Resulting in Relief Valve Lifting & Pipe Vibration,Which Caused Leaks to Develop.Caused by Nearly Closed post-filter Inlet Valve.Repaired 2 Leaks in Line ML20237B7461998-07-22022 July 1998 1998 Defueled Emergency Plan Exercise Scenario Manual, Conducted on 980722 ML20202D1621998-06-30030 June 1998 Safety Evaluation Supporting Amend 193 to License DPR-61 CY-98-068, Follow-up to Verbal Notification on 980413 of Film on Discharge Canal.Investigation Continuing.Samples Collected for Petroleum Analyses & Biological Characterization at Intake Structure & Discharge Canal.Replaced Sorbent Booms1998-04-15015 April 1998 Follow-up to Verbal Notification on 980413 of Film on Discharge Canal.Investigation Continuing.Samples Collected for Petroleum Analyses & Biological Characterization at Intake Structure & Discharge Canal.Replaced Sorbent Booms CY-98-045, Ro:On 980212,0219,0225 & 0312,separate Sheens of Approx One Cup of oil-like Substance Was Observed at Discharge Canal. Cause Has Not Been Clearly Identified.Called in Vendor Spill to Install Sorbent Booms to Absorb Sheen.W/One Drawing1998-04-13013 April 1998 Ro:On 980212,0219,0225 & 0312,separate Sheens of Approx One Cup of oil-like Substance Was Observed at Discharge Canal. Cause Has Not Been Clearly Identified.Called in Vendor Spill to Install Sorbent Booms to Absorb Sheen.W/One Drawing ML20217F0611998-03-31031 March 1998 Historical Review Team Rept ML20217A0001998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Haddam Neck Plant ML20217K2101998-03-27027 March 1998 Safety Evaluation Supporting Amend 192 to License DPR-61 CY-98-046, Follow-up to 980311 Verbal Notification of Film on Discharge Canal.Cause Not Yet Determined.Film Is Contained & Will Be Absorbed by Containment & Sorbent Booms That Were in Place in Discharge Canal1998-03-12012 March 1998 Follow-up to 980311 Verbal Notification of Film on Discharge Canal.Cause Not Yet Determined.Film Is Contained & Will Be Absorbed by Containment & Sorbent Booms That Were in Place in Discharge Canal ML20216D6531998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Haddam Neck Plant ML20217D7381998-02-28028 February 1998 Revised MOR for Feb 1998 Haddam Neck Plant CY-98-012, Monthly Operating Rept for Jan 1998 for Connecticut Yankee Haddam Neck Plant1998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Connecticut Yankee Haddam Neck Plant CY-98-010, Annual Rept for 10CFR50.59,Jan-Dec,19971997-12-31031 December 1997 Annual Rept for 10CFR50.59,Jan-Dec,1997 ML20198N6681997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Haddam Neck Plant ML20217P4861997-12-31031 December 1997 1997 Annual Financial Rept, for Cyap ML20199L5891997-12-24024 December 1997 Independent Analysis & Evaluation of AM-241 & Transuranics & Subsequent Dose to Two Male Workers at Connecticut Yankee Atomic Power Plant ML20203K4271997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for Haddam Neck Plant ML20199B1141997-10-31031 October 1997 Monthly Operating Rept for Oct 1997 for Haddam Neck Plant ML20198M8101997-10-14014 October 1997 SER Accepting Proposed Revs to Util Quality Assurance Program at Facility ML20198J8811997-09-30030 September 1997 Monthly Operating Rept for Sept 1997 for Haddam Neck Plant ML20210P8721997-08-31031 August 1997 Post Decommissioning Activities Rept, for Aug 1997 ML20217Q3171997-08-31031 August 1997 Addl Changes to Proposed Rev 1 to QA Program ML20210U9301997-08-31031 August 1997 Monthly Operating Rept for Aug 1997 for Haddam Neck Plant CY-97-082, Special Rept:On 970708,routine Surveillance Testing of Seismic Monitoring Sys Instrumentation Revealed,Data Was Not Being Reproduced by Portion of Playback Sys.Station Presently Pursuing Replacement of Seismic Monitoring Sys1997-08-14014 August 1997 Special Rept:On 970708,routine Surveillance Testing of Seismic Monitoring Sys Instrumentation Revealed,Data Was Not Being Reproduced by Portion of Playback Sys.Station Presently Pursuing Replacement of Seismic Monitoring Sys ML20210L0521997-07-31031 July 1997 Monthly Operating Rept for July 1997 for HNP ML20149E4451997-06-30030 June 1997 Monthly Operating Rept for June 1997 for Haddam Neck Plant ML20141A0041997-05-31031 May 1997 Independent Assessment of Radiological Controls Program at Cyap Haddam Neck Plant Final Rept May 1997 ML20140H5241997-05-31031 May 1997 Monthly Operating Rept for May 1997 for Haddam Neck Plant ML20141K4201997-05-22022 May 1997 Safety Evaluation Supporting Amend 191 to License DPR-61 ML20141D4141997-04-30030 April 1997 Monthly Operating Rept for Apr 1997 for Connecticut Yankee Haddam Neck ML20138G5901997-04-25025 April 1997 Proposed Rev 1 to Cyap QA Program for Haddam Neck Plant ML20137W8051997-03-31031 March 1997 Monthly Operating Rept for Mar 1997 for Haddam Neck Plant ML20137H3031997-03-31031 March 1997 Rev 2 to Nuclear Training Loit/Lout Audit Reviews ML20137C6281997-03-14014 March 1997 Redacted Version of Rev 1 to Nuclear Training Loit/Lout Audit Reviews ML20137A0801997-02-28028 February 1997 Monthly Operating Rept for Feb 1997 for Haddam Neck Plant ML20135C5101997-02-26026 February 1997 1996 Refuel Outage ISI Summary Rept for CT Yankee Atomic Power Co B16268, Special Rept:On 970205,declared Main Stack-Wide Range Noble Gas Monitor Inoperable.Caused by Inadequate Calibr Methods. Will Revise Calibr Procedure to Technique to Demonstrate Accuracy & Linearity Over Intended Range of Monitor1997-02-19019 February 1997 Special Rept:On 970205,declared Main Stack-Wide Range Noble Gas Monitor Inoperable.Caused by Inadequate Calibr Methods. Will Revise Calibr Procedure to Technique to Demonstrate Accuracy & Linearity Over Intended Range of Monitor ML20135E3221997-02-13013 February 1997 Independent Review Team Rept 1996 MP -1 Lout NRC Exam Failures ML20134L2751997-02-0303 February 1997 Draft Rev to GPRI-30, Spent Fuel Storage Facility Licensing Basis/Design Basis ML20138K5721997-01-31031 January 1997 Monthly Operating Rept for Jan 1997 for Haddam Neck Plant.W/ ML20134L2791997-01-10010 January 1997 Rev 0 to QA Program Grpi ML20134L2911997-01-0808 January 1997 Rev 0 to UFSAR Rev Grpi ML20134L2721996-12-31031 December 1996 Commitment Mgt Grpi 1999-04-28
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Docket No. 50-213 i
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Attachment 1
- Haddam Neck Plant 1
Integrated Safety Assessment Program
] Public Safety Impact Model Project Analyses Summaries i
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, 8606270304 860613 PDR ADOCK 05000213 P PDR .
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, s ISAP #1.27 Compliance with 10CFR50.46 Safety Isste
. Small break loss-of-coolant accidents (LOCAs) are one of several categories of events which, if not successfully mitigated, will lead to core melt. To ensure that each plant has the capability to provide coolant after a small break LOCA, redundant high pressure injection pumps nust be provided as part of the plant's emergency core cooling system (ECCS). Connecticut Yankee has two diverse means of providing make-up after a small break LOCA. Two 360-gpm centrifugal charging pumps are part of the Chemical and Volume Control System (CVCS). One of these pumps is continually running during normal operation and the standby pump will start on a safety injection (SI) signal if offsite power is available. Two High Pressure Safety Injection (HPSI) pumps which are part of the CY ECCS are also available for high pressure coolant injection. These pumps automatically start on a SI signal, regardless of the availability of offsite power.
To ensure that these redundant and diverse means of high pressure injection are adequate to provide sufficient make-up for all small break LOCAs in all locations, plant-specific calculations are necessary to show that CY has adequate capability to prevent core melt in the event of a small break LOCA.
Proposed Project 10CFR50.46 requires each light-water reactor to be equipped with an ECCS which can provide cooling for all postulated small break LOCAs in the reactor coolant syster (RCS). In addition, plant-specific calculations using NRC-approved models for small break LOCAs are also required to show that the ECCS is capable of mitigating small break LOCAs. The proposed project is an evaluation of the CY ECCS including calculations to show that the ECCS is in compliance with 10CFR50.46. In addition, the proposed project would include any necessary modifications to the ECCS which may be needed to ensure that all small break LOCAs can be successfully mitigated.
CONNECTICUT YANKEE INTEGRATED SAFETY ASSESSMENT PROGRAM J
Analysis of Public Safety Impact Northeast Utilities has performed plant-specific calculations which show that the CY ECCS is in compliance with 10CFR50.46 and is capable of mitigating all small break LOCAs in the injection phase (Reference 1). These calculations only credited the HPSI pumps and did not include the charging pumps. The Best Estimate LOCA Analysis for CY (Reference 2) showed that one HPSI or one charging pump is adequate for all small break LOCAs for both the injection and recirculation phases except for a certain range of break sizes (0.02 to 0.045 2
ft ) in the RCS loop 2 cold leg and for breaks of less than 0.038 ft in the charging line. For these breaks the ECCS was not capable of providing core cooling in the high pressure recirculation mode. At the time the CY Probabilistic Safety Study (PSS) was performed, only the charging pumps were able to provide high pressure recirculation and the only injection path was via the loop 2 cold leg. Since that time modifications have been made to enable l the HPSI pumps to provide high-pressure recirculation in the event that the l charging pumps are unavailable or inadequate (i.e., the small break LOCA cccurs in the loop 2 cold leg). These modifications allow the operator to align the ;
HPSI pumps to take suction from the residual heat removal (RHR) pumps during l the recirculation mode and inject coolant into the cold legs. This method of recirculation is not single-failure proof, however.
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A sensitivity analysis was conducted to determine the effect of using the HPSI ,
l pumps for high pressure recirculation on the core melt frequency. This 1 analysis incorporated a postulated single failure-proof high pressure l recirculation system with the HPSI pumps taking suction from the RHR punns as a j back-up method to using the charging pumps taking suction from the RHR pumps I for recirculation. New fault trees where constructed and quantified for the l new high pressure recirculation method and event trees for small- and medium-break LOCAs were revised and requantified to determine a new core melt ,
1 frequency. !
l With respect to meeting the plant-specific small break LOCA calculations j required by 10CFR50.46, Northeast Utilities has performed these calculations for the injection phase of the small break LOCA using procedural and analytical methods which conform to the criteria set forth in 10CFR50.46 and Appendix K (to the extent it is applicable recognizing that stainless steel cladding is i
CONNECTIClTr YANKEE I E EGRATED SAFETY ASSESSME E PROGRAM
, 6 utilized for all but four fuel assemblies) of 10CFR50 (Reference 1). The CY Best Estimate Analysis shows that the ECCS is adequate for all small break j LOCAs in the recirculation phase except for the limited range of breaks previously described. This analysis used the same methods and computer code
! which were employed in the analysis documented in Reference 1 with some minor changes. These changes only affect the analysis of the injection phase. Since there would be no significant difference between the best estimate analysis and an analysis which meets the criteria specified in 10CFR50.46 for the recirculation mode, further analyses of the recirculation mode for small break l LOCAs to show compliance with 10CFR50.46 are unwarranted and would have no 1
impact on public safety.
Results The sensitivity analysis which assumes a single-failure proof high pressure recirculation method using the HPSI pumps as a back-up method to using the charging pumps for recirculation shows a decrease in the core melt frequency of 1.57 x 10-4, a 28.6% decrease from the present core melt frequency of 5.48 x 10 . This decrease in core melt frequency is larger than the 14% of the core melt frequency which is contribreed by the small break LOCAs in the RCS loop 2 cold leg because using the FPSI pumps as a back-up to the charging pumps increases the reliability of providing recirculation for all small- and medium-break LOCAs.
All of the core melt sequences affected by this change are in consequence category 5 (mean consequence: 2.8 x 103 man-rem). The resulting public risk is calculated as follows:
R =Tx AP1 xC 1 where h = total change in public risk, man-rems T = remaining plant life, 20 years i AP 1
= change in core melt frequency, 1.57 x 10-N/ year i l
l CONNECTICUT YANKEE INTEGRATED SAFETY ASSESSMENT PROGRAM
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= offsite public consequences associated with category l 3
5 - 2.8'x 10 man-rems l R = (20 years) x (1.57 x 10# year-l)x (2.8 x 103 ) j
~ 9 man-rems l l
Based on the Public Safety. Impact Model, this reduction in risk is equivalent !
to a rank of 0.2 on a scale of -10 to +10.
References
- 1. W.G. Counsil letter to John A. Zwolinski, "Haddam Neck Plant Small Break I
LOCA Topical Report, TMI Action Plan Items II.K.3.5, II.K.3.30, and II.K.3.31," Docket No. 50-213, dated December 19, 1984.
- 2. J.F. Opeka letter to Christopher I. Grimes, "Haddam Neck Plant Probabilistic Safety Study - Summary Report and Results," Docket No.
50-213, dated March 31, 1986.
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CONNECTICUT YANKEE INTEGRATED SAFETY ASSESSMENT PROGRAM l
ISAP #2.06 Evaluation of RCS Loop Isolation Valves to Mitigate Steam Generator Tube Rupture Safety Issue The RCS Loop Isolation Valves can be used to isolate a primary loop in cases of RCS pump seal failure or other problems associated with one loop. Each primary loop is equipped with two motor-operated isolation valves at either side of the reactor coolant pump (RCP) and the steam generator. This project was initiated to evaluate the acceptability of utilizing the RCS loop isolation valves (LIV's) to mitigate the consequences of a steam generator tube rupture (SGTR) event, particularly where a steam generator safety relief valve is stucl: open.
The LIV's at CY are capable of closing against a maximum differential pressure of 500 psi. CY transient analyses have shown that only in situations where the high pressure steam dump system is unavailable might this limit of 500 psi be exceeded following reactor trip. (The possibility exists of opening the pressurizer PORVs but this was not considered here). High pressure steam dump to condenser will eventually be unavailable following such events as loss of offsite power (LOSP), loss of condenser vacuum or loss of control air. The high pressure steam dump system consists of ten air-operated valves (A0Vs).
Proposed Project l
The proposed project concerns the qualification of the valve and its operator to function in the presence of a possible pressure drop across the valve disk.
Since LIV's take about three minutes to close, a pressure drop may develop when the valves are nearly closed and the steam generator side of the valves begins i
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. to depressurize due to the leakage into the generator. (It should be noted ,
that credit for closure of the LIV for non-LOSP events was taken in the CY PSS for cases where a safety relief valve on the faulted steam generator is stuck open). This analysis evaluates the effect of a failure of the LIV's to close. ;
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CONNECTICUT YANKEE INTEGRATED SAFETY ASSESSMENT PROGRAM
l Analysis of Public Safety Impact The dominant contributor to the unavailability of steam dump to condenser is a LOSP. Loss of air or other causes of failure of the A0Vs are relatively small. There is no reason to believe that there is any correlation between a SGIR event and a LOSP, or any of the other failures. They can therefore be treated independently. A SGTR event followed by LOSP for about one hour has a frequency of [ Reference 1]:
1.7E-2 yr-I x (.17 yr /(365 days /yr x 24 hr/dy) x 1 hr) x 0.5
= 1.7E-7 yr .
where a factor of 0.5 is conservatively applied for probability of not recovering offsite power in one hour. The period of one hour is chosen conservatively, since the amount of time required for depressurization of the primary system is only a few minutes. This frequency is then multiplied by the probabilities of various sequences leading to core melt (plant damage state V2) {
that involve the availability of the loop isolation valves. The probability of failing to utilize the loop isolation valves is set equal to one in all of these sequences.
Results The frequency of the V2 plar.t damage state is decreased by 1.0E-9 yr-l if the LIV could be used under all conditions of SGTR coincident with loss of high pressure steam dump capability. The decrease in public risk is given by:
R = 1.0E-9 yr-l x 1.6E+6 man-rem x 20 yr = 3.2E-2 man-rem This project has an ISAP score of less than 0.001.
References
- 1. J.F. Opeka letter to Christopher I. Grimes, "Haddam Neck Plant Probabilistic Safety Study -
Summary Report and Results," Docket No.
50-213, dated March 31, 1986.
CONNECTICUT YANKEE INTEGRATED SAFETY ASSESSMENT PROGRAM