ML18094B148: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
Line 17: Line 17:


=Text=
=Text=
{{#Wiki_filter:,.. Public Service Electric and Gas Company Steven E. Miltenberger Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4199 ) Vice President and Chief Nuclear Officer NOV 0 1 1989 NLR-N89218 U.S. Nuclear Regulatory Commission Document Control **Desk Washington, D. c. 20555 Gentlemen:
{{#Wiki_filter:,..
Public Service Electric and Gas Company Steven E. Miltenberger                   Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4199
                                                                                                                      )
Vice President and Chief Nuclear Officer NOV 0 1 1989 NLR-N89218 U.S. Nuclear Regulatory Commission Document Control **Desk Washington, D. c. 20555 Gentlemen:
INDIVIDUAL PLANT EXAMINATION FOR SEVERE ACCIDENT VULNERABILITIES NRC GENERIC LETTER 88-20 AND GENERIC LETTER 88-20, SUPPLEMENT 1 SALEM AND HOPE CREEK GENERATING STATIONS DOCKET NOS. 50-272, 50-311 AND 50-354 Public Service Electric and Gas Company (PSE&G) hereby provides a response to your November 23, 1988 and August 29, 1989 letters requesting a description of the method to be used for performing the Individual Plant Examination (IPE), including milestones and schedules.
INDIVIDUAL PLANT EXAMINATION FOR SEVERE ACCIDENT VULNERABILITIES NRC GENERIC LETTER 88-20 AND GENERIC LETTER 88-20, SUPPLEMENT 1 SALEM AND HOPE CREEK GENERATING STATIONS DOCKET NOS. 50-272, 50-311 AND 50-354 Public Service Electric and Gas Company (PSE&G) hereby provides a response to your November 23, 1988 and August 29, 1989 letters requesting a description of the method to be used for performing the Individual Plant Examination (IPE), including milestones and schedules.
PSE&G initiated the PRA Program in advance of Generic Letter 88-20 issuance.
PSE&G initiated the PRA Program in advance of Generic Letter 88-20 issuance. It was our desire to produce a risk based tool that could be used during the plant modification process and aid in the prioritization of resources.
It was our desire to produce a risk based tool that could be used during the plant modification process and aid in the prioritization of resources.
It should be noted that, although we refer to other plant analyses, it is our intent to do a largely plant-specific analysis for Salem and Hope Creek.
It should be noted that, although we refer to other plant analyses, it is our intent to do a largely plant-specific analysis for Salem and Hope Creek. The resources required to accomplish this work is estimated at 18 manyears, of which about 60% is PSE&G and 40% is consultant support. Please contact us if you have any questions regarding this transmittal.
The resources required to accomplish this work is estimated at 18 manyears, of which about 60% is PSE&G and 40% is consultant support.
Enclosure ( 8911070400 891101 PDR ADOCK 05000272 p PNU \ \ Sincerely, Document Control Desk NLR-N89218 C Mr. J. C. Stone Licensing Project Manager -Salem Ms. K. Halvey Gibson Senior Resident Inspector  
Please contact us if you have any questions regarding this transmittal.
-Salem Mr. c. Y. Shiraki Licensing Project Manager -Hope Creek Mr. G. w. Meyer Senior Resident Inspector  
Sincerely, Enclosure
-Hope Creek Mr. w. T. Russell, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, N. J. 08625 NOV 0 1 1989 REF: NLR-N89218 STATE OF NEW JERSEY ) ) SS. COUNTY OF SALEM ) s. E. Miltenberger, being duly sworn according to law deposes and says: I am Vice President and Chief Nuclear Officer of Public Service Electric and Gas Company, and as such, I find the matters set forth in our letter dated NOV 0 1 19S9 , concerning the Salem Unit Nos. 1 and 2 and Hope Creek Generating stations, are true to the best of my knowledge, information and belief. Subscribed and Sworn to before me . 'Jf f" day of 1989 .. / .* -
                                                                    \
.** *1ic.1tary
( 8911070400 891101                                   \
__ Public of New Jersey ENCLOSURE RESPONSE TO NRC GENERIC LETTER 88-20 AND SUPPLEMENT 1 SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311 NLR-N89218 Public Service Electric And Gas Company (PSE&G) is developing a program of Probabilistic Risk Assessment (PRA) for application to the Salem Generating Station (SGS), and Hope Creek Generating station (HCGS). The completed program will meet the requirements and objectives for Individual Plant Examination (IPE), as specified in Generic Letter 88-20 and the supplemented guidance of NUREG-1335.
PDR ADOCK 05000272 p                           PNU
Through the performance of the IPE, PSE&G will fulfill the stated objectives, namely: (1) developing an appreciation of severe accident behavior; (2) understanding the most likely severe accident sequences; (3) gaining a more quantitative understanding of the overall probabilities of core damage and fission product releases; and (4) if necessary, reducing the overall probabilities of core damage and fission product release by modifying, where appropriate, hardware and procedures that would assist in preventing or mitigating severe accidents.
 
This letter summarizes the current status of the PRA program at PSE&G and: 1. Identifies the method and approach selected for IPE performance, 2. Describes the method to be used, and 3. Identifies the schedule and milestones for IPE performance and subsequent NRC reportability.
Document Control Desk                           NOV 0 1 1989 NLR-N89218 C   Mr. J. C. Stone Licensing Project Manager - Salem Ms. K. Halvey Gibson Senior Resident Inspector - Salem Mr. c. Y. Shiraki Licensing Project Manager - Hope Creek Mr. G. w. Meyer Senior Resident Inspector - Hope Creek Mr. w. T. Russell, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, N. J. 08625
PSE&G has recently completed a Level 1 PRA for SGS in the context of NUREG/CR-2300, including internal flooding.initiating events. The model utilizes a support state approach characterized by large fault trees and small event trees. Comprehensive dependency matrices were developed and utilized in the fault tree system modeling, to provide assurance that all relevant dependencies were considered.
 
The Level 1 PRA sequences have been characterized in terms of their plant damage state and functional failure type. This grouping provides for easy transfer of sequence information to the Containment Analysis and helps provide a basis for development of a subsequent Accident Management Program. The effort to date has been divided approximately equally between PSE&G and consultants.
REF: NLR-N89218 STATE OF NEW JERSEY             )
Future plans call for reduced consultant participation, because it is PSE&G's intention to produce a risk assessment tool that is essentially free of consultant support. Page 1 of 4 NLR-N89218 The aforementioned Level 1 PRA forms the basic tool for the front end of the IPE. Some modifications and additions must still be made to the model and its associated report, before it can be considered the documented basis for the IPE. The internal flooding modeling for SGS was performed upon completion of all other aspects of the Level 1 analysis.
                                    ) SS.
The flooding assessment thus benefited from the detailed plant model developed through the remainder of the Level 1 evaluation.
COUNTY OF SALEM                 )
The flooding analysis considered:
: s. E. Miltenberger, being duly sworn according to law deposes and says:
potential flooding sources, flooding pathways to key components, the probability of flooding, the potential for flood induced plant accidents or transients, the likelihood of system failures due to flooding, and the ability to detect and mitigate postulated flooding conditions.
I am Vice President and Chief Nuclear Officer of Public Service Electric and Gas Company, and as such, I find the matters set forth in our letter dated           NOV 0 1 19S9   , concerning the Salem Unit Nos. 1 and 2 and Hope Creek Generating stations, are true to the best of my knowledge, information and belief.
A similar approach has been adopted for the HCGS study. The SGS and HCGS containment evaluations will include a plant-specific containment event tree, consistent with Section 2.2.2.5 of NUREG-1335.
Subscribed and Sworn to before me thi~
The potential containment failure mechanisms described in Table 2.2 of NUREG-1335 will be addressed in these evaluations.
      .. /
The approach is to use the results of similar plant evaluations and plant-specific analysis, as required.
            . 'Jf f"   day of Lj}6JH/rfv~    1989
An evaluation of the principal differences between the PSE&G facilities and any reference facilities will be performed, to ensure that the reference calculations and results are directly applicable or to identify the impact of any differences.
.* -~-l~LrL~ALtU-Z
.** *1ic.1tary__ Public of New Jersey
 
NLR-N89218 ENCLOSURE RESPONSE TO NRC GENERIC LETTER 88-20 AND SUPPLEMENT 1 SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311 Public Service Electric And Gas Company (PSE&G) is developing a program of Probabilistic Risk Assessment (PRA) for application to the Salem Generating Station (SGS), and Hope Creek Generating station (HCGS). The completed program will meet the requirements and objectives for Individual Plant Examination (IPE), as specified in Generic Letter 88-20 and the supplemented guidance of NUREG-1335.
Through the performance of the IPE, PSE&G will fulfill the stated objectives, namely:   (1) developing an appreciation of severe accident behavior; (2) understanding the most likely severe accident sequences; (3) gaining a more quantitative understanding of the overall probabilities of core damage and fission product releases; and (4) if necessary, reducing the overall probabilities of core damage and fission product release by modifying, where appropriate, hardware and procedures that would assist in preventing or mitigating severe accidents.
This letter summarizes the current status of the PRA program at PSE&G and:
: 1. Identifies the method and approach selected for IPE performance,
: 2. Describes the method to be used, and
: 3. Identifies the schedule and milestones for IPE performance and subsequent NRC reportability.
PSE&G has recently completed a Level 1 PRA for SGS in the context of NUREG/CR-2300, including internal flooding.initiating events.
The model utilizes a support state approach characterized by large fault trees and small event trees. Comprehensive dependency matrices were developed and utilized in the fault tree system modeling, to provide assurance that all relevant dependencies were considered. The Level 1 PRA sequences have been characterized in terms of their plant damage state and functional failure type. This grouping provides for easy transfer of sequence information to the Containment Analysis and helps provide a basis for development of a subsequent Accident Management Program.
The effort to date has been divided approximately equally between PSE&G and consultants. Future plans call for reduced consultant participation, because it is PSE&G's intention to produce a risk assessment tool that is essentially free of consultant support.
Page 1 of 4
 
NLR-N89218 The aforementioned Level 1 PRA forms the basic tool for the front end of the IPE. Some modifications and additions must still be made to the model and its associated report, before it can be considered the documented basis for the IPE.
The internal flooding modeling for SGS was performed upon completion of all other aspects of the Level 1 analysis. The flooding assessment thus benefited from the detailed plant model developed through the remainder of the Level 1 evaluation. The flooding analysis considered: potential flooding sources, flooding pathways to key components, the probability of flooding, the potential for flood induced plant accidents or transients, the likelihood of system failures due to flooding, and the ability to detect and mitigate postulated flooding conditions.
A similar approach has been adopted for the HCGS study.
The SGS and HCGS containment evaluations will include a plant-specific containment event tree, consistent with Section 2.2.2.5 of NUREG-1335. The potential containment failure mechanisms described in Table 2.2 of NUREG-1335 will be addressed in these evaluations. The approach is to use the results of similar plant evaluations and plant-specific analysis, as required. An evaluation of the principal differences between the PSE&G facilities and any reference facilities will be performed, to ensure that the reference calculations and results are directly applicable or to identify the impact of any differences.
The principal elements of the containment assessment are described below:
The principal elements of the containment assessment are described below:
* Potential vulnerabilities identified through plant walkdowns or design evaluations will be assessed to determine their impact on containment performance.
* Potential vulnerabilities identified through plant walkdowns or design evaluations will be assessed to determine their impact on containment performance.
* Containment ultimate capability will be determined by performing plant-specific engineering analyses or evaluating containment performance evaluations performed for similar facilities.
* Containment ultimate capability will be determined by performing plant-specific engineering analyses or evaluating containment performance evaluations performed for similar facilities. Identified ultimate capability controlling features will be compared to the PSE&G facilities, to determine the degree of applicability of these assessments to the plant-specific evaluations.
Identified ultimate capability controlling features will be compared to the PSE&G facilities, to determine the degree of applicability of these assessments to the plant-specific evaluations.
* A containment event tree will be developed for each assessment. The event trees will include sufficient detail to reflect the effects of plant systems and emergency procedure changes, yet concise enough to effectively communicate the results. The containment event tree will be analyzed for each plant damage state. Success paths for recovery from degraded core conditions within the reactor vessel will be included.
* A containment event tree will be developed for each assessment.
Emergency procedures and operator recovery actions taken during severe accident progression (up to the point of postulated containment failure) and system recovery/repair, will be incorporated into the Page 2 of 4
The event trees will include sufficient detail to reflect the effects of plant systems and emergency procedure changes, yet concise enough to effectively communicate the results. The containment event tree will be analyzed for each plant damage state. Success paths for recovery from degraded core conditions within the reactor vessel will be included.
 
Emergency procedures and operator recovery actions taken during severe accident progression (up to the point of postulated containment failure) and system recovery/repair, will be incorporated into the Page 2 of 4
*
*
* u
* u analysis.
* analysis.
* This process will result in a more realistic NLR-N89218 assessment, including insights for accident management evaluation. The nodes of the containment event tree will be quantified, and the release scenarios identified characterized in terms of timing and severity level.
This process will result in a more realistic assessment, including insights for accident management evaluation.
The nodes of the containment event tree will be quantified, and the release scenarios identified characterized in terms of timing and severity level. NLR-N89218
* The timing and progression of identified sequences will be assessed based upon plant specific or representative calculations from plants similar to SGS and HCGS.
* The timing and progression of identified sequences will be assessed based upon plant specific or representative calculations from plants similar to SGS and HCGS.
* A series of source term code results will be compiled to characterize the severity of sequences postulated through the containment event tree formation.
* A series of source term code results will be compiled to characterize the severity of sequences postulated through the containment event tree formation. The impact of pathway, magnitude of release, rate of release and mitigation effectiveness will be included in the characterization of each containment event tree release path. Similar sequences will be combined to develop a comprehensive set of release likelihoods. If deemed necessary, direct plant specific source terms will be calculated for selected sequences.
The impact of pathway, magnitude of release, rate of release and mitigation effectiveness will be included in the characterization of each containment event tree release path. Similar sequences will be combined to develop a comprehensive set of release likelihoods.
* Uncertainties will be addressed by sensitivity analyses, comparisons between results obtained from different containment performance codes employing different approaches to describing complex phenomena, or a combination of both. Emphasis will be placed on uncertainties in phenomena which directly impact accident management considerations. Less emphasis will be given to phenomena which might affect the source term, but do not strongly impact the selection of strategies for coping with severe accidents.
If deemed necessary, direct plant specific source terms will be calculated for selected sequences.
Upon completion of the Level 1 and containment performance assessments, an evaluation of vulnerabilities and insights will be performed for the SGS and HCGS. These evaluations will consider those sequences meeting the screening criteria defined in Generic Letter 88-20, and include the consideration of uncertainties to enhance the insights developed. Selected USI and GSI topics may be included, but as a minimum, USI A-45 will be resolved. USI and GSI topics will be evaluated in terms of the PRA insights and a summary of plant vulnerability for each issue developed.
* Uncertainties will be addressed by sensitivity analyses, comparisons between results obtained from different containment performance codes employing different approaches to describing complex phenomena, or a combination of both. Emphasis will be placed on uncertainties in phenomena which directly impact accident management considerations.
PSE&G will evaluate each potential accident vulnerability identified.through the performance of these assessments. Any identified-vulnerabilities that warrant correction will be thoroughly evaluated for potential improvements. Consideration will be given to the proposed Mark I improvements described in Enclosure 2 of Generic Letter Supplement 1, during the analysis for HCGS. The results of the IPE will be submitted to the NRC using a two (2) tier approach. The information will include the results of the examinations and a summary of the insights gained.
Less emphasis will be given to phenomena which might affect the source term, but do not strongly impact the selection of strategies for coping with severe accidents.
Detailed documentation will be developed in a traceable manner Page 3 of 4
Upon completion of the Level 1 and containment performance assessments, an evaluation of vulnerabilities and insights will be performed for the SGS and HCGS. These evaluations will consider those sequences meeting the screening criteria defined in Generic Letter 88-20, and include the consideration of uncertainties to enhance the insights developed.
 
Selected USI and GSI topics may be included, but as a minimum, USI A-45 will be resolved.
    . '~
USI and GSI topics will be evaluated in terms of the PRA insights and a summary of plant vulnerability for each issue developed.
                      *
PSE&G will evaluate each potential accident vulnerability identified.through the performance of these assessments.
* NLR-N89218 and retained by PSE&G for use in the ongoing risk evaluation and safety assessment program. The scope of information transmitted to the NRC will be consistent with the general guidance in Table 2.1 of NUREG-1335.
Any identified-vulnerabilities that warrant correction will be thoroughly evaluated for potential improvements.
Generic Letter 88-20 states in part, "*** the quality and comprehensiveness of the results derived from an IPE depend on
Consideration will be given to the proposed Mark I improvements described in Enclosure 2 of Generic Letter Supplement 1, during the analysis for HCGS. The results of the IPE will be submitted to the NRC using a two (2) tier approach.
'
The information will include the results of the examinations and a summary of the insights gained. Detailed documentation will be developed in a traceable manner Page 3 of 4
0 the vigor with which the analyst applies the method of examination and on the utility's commitment to the intent of the IPE." PSE&G initiated the current PRA program based upon the desire to: evaluate its facilities, produce a risk based tool to assist in the plant modification process and aid in the prioritization of resources. The program is being developed by a dedicated PSE&G PRA group with the assistance of station personnel and consultants. Operations Department personnel and System Engineers review results for consistency with plant design and operation. Thus, the risk evaluation effort is the product of PSE&G.
* * . NLR-N89218 and retained by PSE&G for use in the ongoing risk evaluation and safety assessment program. The scope of information transmitted to the NRC will be consistent with the general guidance in Table 2.1 of NUREG-1335.
The program is designed to proceed at a pace that supports active and thorough involvement by all appropriate personnel, and to allow parallel utilization of the current model for risk assessment activities independent of the IPE. Our schedule is consistent with this level of involvement in PRA development and the associated evaluation of any resulting insights.
Generic Letter 88-20 states in part, "*** the quality and comprehensiveness of the results derived from an IPE depend on '0 the vigor with which the analyst applies the method of examination and on the utility's commitment to the intent of the IPE." PSE&G initiated the current PRA program based upon the desire to: evaluate its facilities, produce a risk based tool to assist in the plant modification process and aid in the prioritization of resources.
* Completion of HCGS Level 1 analysis. - June 1990
The program is being developed by a dedicated PSE&G PRA group with the assistance of station personnel and consultants.
* Completion of enhancements to the       - Nov. 1990 SGS Level 1 analysis.
Operations Department personnel and System Engineers review results for consistency with plant design and operation.
* Completion of HCGS containment - Oct. 1991 performance analysis.
Thus, the risk evaluation effort is the product of PSE&G. The program is designed to proceed at a pace that supports active and thorough involvement by all appropriate personnel, and to allow parallel utilization of the current model for risk assessment activities independent of the IPE. Our schedule is consistent with this level of involvement in PRA development and the associated evaluation of any resulting insights.
* Completion of SGS containment   - Oct. 1992 performance analysis.
* Completion of HCGS Level 1 analysis.  
* Completion of final assessments   - July 1993 and recommendations, submittal
-June 1990
              - of the IPE report.
* Completion of enhancements to the SGS Level 1 analysis.  
PSE&G will provide a brief summary report to the NRC upon completion of each milestone.
-Nov. 1990
* Completion of HCGS containment  
-Oct. 1991 performance analysis.
* Completion of SGS containment  
-Oct. 1992 performance analysis.
* Completion of final assessments  
-July 1993 and recommendations, submittal  
-of the IPE report. PSE&G will provide a brief summary report to the NRC upon completion of each milestone.
Page 4 of 4}}
Page 4 of 4}}

Revision as of 11:05, 21 October 2019

Responds to Generic Ltr 88-20, Individual Plant Exam for Severe Accident Vulnerabilities, & 881123 & 890829 Ltrs Requesting Description of Method to Be Used for Performing Individual Plant Exam
ML18094B148
Person / Time
Site: Salem, Hope Creek, 05000000
Issue date: 11/01/1989
From: Miltenberger S
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-1335 GL-88-20, NLR-N89218, NUDOCS 8911070400
Download: ML18094B148 (7)
v • d • e


Text

,..

Public Service Electric and Gas Company Steven E. Miltenberger Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4199

)

Vice President and Chief Nuclear Officer NOV 0 1 1989 NLR-N89218 U.S. Nuclear Regulatory Commission Document Control **Desk Washington, D. c. 20555 Gentlemen:

INDIVIDUAL PLANT EXAMINATION FOR SEVERE ACCIDENT VULNERABILITIES NRC GENERIC LETTER 88-20 AND GENERIC LETTER 88-20, SUPPLEMENT 1 SALEM AND HOPE CREEK GENERATING STATIONS DOCKET NOS. 50-272, 50-311 AND 50-354 Public Service Electric and Gas Company (PSE&G) hereby provides a response to your November 23, 1988 and August 29, 1989 letters requesting a description of the method to be used for performing the Individual Plant Examination (IPE), including milestones and schedules.

PSE&G initiated the PRA Program in advance of Generic Letter 88-20 issuance. It was our desire to produce a risk based tool that could be used during the plant modification process and aid in the prioritization of resources.

It should be noted that, although we refer to other plant analyses, it is our intent to do a largely plant-specific analysis for Salem and Hope Creek.

The resources required to accomplish this work is estimated at 18 manyears, of which about 60% is PSE&G and 40% is consultant support.

Please contact us if you have any questions regarding this transmittal.

Sincerely, Enclosure

\

( 8911070400 891101 \

PDR ADOCK 05000272 p PNU

Document Control Desk NOV 0 1 1989 NLR-N89218 C Mr. J. C. Stone Licensing Project Manager - Salem Ms. K. Halvey Gibson Senior Resident Inspector - Salem Mr. c. Y. Shiraki Licensing Project Manager - Hope Creek Mr. G. w. Meyer Senior Resident Inspector - Hope Creek Mr. w. T. Russell, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, N. J. 08625

REF: NLR-N89218 STATE OF NEW JERSEY )

) SS.

COUNTY OF SALEM )

s. E. Miltenberger, being duly sworn according to law deposes and says:

I am Vice President and Chief Nuclear Officer of Public Service Electric and Gas Company, and as such, I find the matters set forth in our letter dated NOV 0 1 19S9 , concerning the Salem Unit Nos. 1 and 2 and Hope Creek Generating stations, are true to the best of my knowledge, information and belief.

Subscribed and Sworn to before me thi~

.. /

. 'Jf f" day of Lj}6JH/rfv~ 1989

.* -~-l~LrL~ALtU-Z

.** *1ic.1tary__ Public of New Jersey

NLR-N89218 ENCLOSURE RESPONSE TO NRC GENERIC LETTER 88-20 AND SUPPLEMENT 1 SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311 Public Service Electric And Gas Company (PSE&G) is developing a program of Probabilistic Risk Assessment (PRA) for application to the Salem Generating Station (SGS), and Hope Creek Generating station (HCGS). The completed program will meet the requirements and objectives for Individual Plant Examination (IPE), as specified in Generic Letter 88-20 and the supplemented guidance of NUREG-1335.

Through the performance of the IPE, PSE&G will fulfill the stated objectives, namely: (1) developing an appreciation of severe accident behavior; (2) understanding the most likely severe accident sequences; (3) gaining a more quantitative understanding of the overall probabilities of core damage and fission product releases; and (4) if necessary, reducing the overall probabilities of core damage and fission product release by modifying, where appropriate, hardware and procedures that would assist in preventing or mitigating severe accidents.

This letter summarizes the current status of the PRA program at PSE&G and:

1. Identifies the method and approach selected for IPE performance,
2. Describes the method to be used, and
3. Identifies the schedule and milestones for IPE performance and subsequent NRC reportability.

PSE&G has recently completed a Level 1 PRA for SGS in the context of NUREG/CR-2300, including internal flooding.initiating events.

The model utilizes a support state approach characterized by large fault trees and small event trees. Comprehensive dependency matrices were developed and utilized in the fault tree system modeling, to provide assurance that all relevant dependencies were considered. The Level 1 PRA sequences have been characterized in terms of their plant damage state and functional failure type. This grouping provides for easy transfer of sequence information to the Containment Analysis and helps provide a basis for development of a subsequent Accident Management Program.

The effort to date has been divided approximately equally between PSE&G and consultants. Future plans call for reduced consultant participation, because it is PSE&G's intention to produce a risk assessment tool that is essentially free of consultant support.

Page 1 of 4

NLR-N89218 The aforementioned Level 1 PRA forms the basic tool for the front end of the IPE. Some modifications and additions must still be made to the model and its associated report, before it can be considered the documented basis for the IPE.

The internal flooding modeling for SGS was performed upon completion of all other aspects of the Level 1 analysis. The flooding assessment thus benefited from the detailed plant model developed through the remainder of the Level 1 evaluation. The flooding analysis considered: potential flooding sources, flooding pathways to key components, the probability of flooding, the potential for flood induced plant accidents or transients, the likelihood of system failures due to flooding, and the ability to detect and mitigate postulated flooding conditions.

A similar approach has been adopted for the HCGS study.

The SGS and HCGS containment evaluations will include a plant-specific containment event tree, consistent with Section 2.2.2.5 of NUREG-1335. The potential containment failure mechanisms described in Table 2.2 of NUREG-1335 will be addressed in these evaluations. The approach is to use the results of similar plant evaluations and plant-specific analysis, as required. An evaluation of the principal differences between the PSE&G facilities and any reference facilities will be performed, to ensure that the reference calculations and results are directly applicable or to identify the impact of any differences.

The principal elements of the containment assessment are described below:

  • Potential vulnerabilities identified through plant walkdowns or design evaluations will be assessed to determine their impact on containment performance.
  • Containment ultimate capability will be determined by performing plant-specific engineering analyses or evaluating containment performance evaluations performed for similar facilities. Identified ultimate capability controlling features will be compared to the PSE&G facilities, to determine the degree of applicability of these assessments to the plant-specific evaluations.
  • A containment event tree will be developed for each assessment. The event trees will include sufficient detail to reflect the effects of plant systems and emergency procedure changes, yet concise enough to effectively communicate the results. The containment event tree will be analyzed for each plant damage state. Success paths for recovery from degraded core conditions within the reactor vessel will be included.

Emergency procedures and operator recovery actions taken during severe accident progression (up to the point of postulated containment failure) and system recovery/repair, will be incorporated into the Page 2 of 4

  • u analysis.
  • This process will result in a more realistic NLR-N89218 assessment, including insights for accident management evaluation. The nodes of the containment event tree will be quantified, and the release scenarios identified characterized in terms of timing and severity level.
  • The timing and progression of identified sequences will be assessed based upon plant specific or representative calculations from plants similar to SGS and HCGS.
  • A series of source term code results will be compiled to characterize the severity of sequences postulated through the containment event tree formation. The impact of pathway, magnitude of release, rate of release and mitigation effectiveness will be included in the characterization of each containment event tree release path. Similar sequences will be combined to develop a comprehensive set of release likelihoods. If deemed necessary, direct plant specific source terms will be calculated for selected sequences.
  • Uncertainties will be addressed by sensitivity analyses, comparisons between results obtained from different containment performance codes employing different approaches to describing complex phenomena, or a combination of both. Emphasis will be placed on uncertainties in phenomena which directly impact accident management considerations. Less emphasis will be given to phenomena which might affect the source term, but do not strongly impact the selection of strategies for coping with severe accidents.

Upon completion of the Level 1 and containment performance assessments, an evaluation of vulnerabilities and insights will be performed for the SGS and HCGS. These evaluations will consider those sequences meeting the screening criteria defined in Generic Letter 88-20, and include the consideration of uncertainties to enhance the insights developed. Selected USI and GSI topics may be included, but as a minimum, USI A-45 will be resolved. USI and GSI topics will be evaluated in terms of the PRA insights and a summary of plant vulnerability for each issue developed.

PSE&G will evaluate each potential accident vulnerability identified.through the performance of these assessments. Any identified-vulnerabilities that warrant correction will be thoroughly evaluated for potential improvements. Consideration will be given to the proposed Mark I improvements described in Enclosure 2 of Generic Letter Supplement 1, during the analysis for HCGS. The results of the IPE will be submitted to the NRC using a two (2) tier approach. The information will include the results of the examinations and a summary of the insights gained.

Detailed documentation will be developed in a traceable manner Page 3 of 4

. '~

  • NLR-N89218 and retained by PSE&G for use in the ongoing risk evaluation and safety assessment program. The scope of information transmitted to the NRC will be consistent with the general guidance in Table 2.1 of NUREG-1335.

Generic Letter 88-20 states in part, "*** the quality and comprehensiveness of the results derived from an IPE depend on

'

0 the vigor with which the analyst applies the method of examination and on the utility's commitment to the intent of the IPE." PSE&G initiated the current PRA program based upon the desire to: evaluate its facilities, produce a risk based tool to assist in the plant modification process and aid in the prioritization of resources. The program is being developed by a dedicated PSE&G PRA group with the assistance of station personnel and consultants. Operations Department personnel and System Engineers review results for consistency with plant design and operation. Thus, the risk evaluation effort is the product of PSE&G.

The program is designed to proceed at a pace that supports active and thorough involvement by all appropriate personnel, and to allow parallel utilization of the current model for risk assessment activities independent of the IPE. Our schedule is consistent with this level of involvement in PRA development and the associated evaluation of any resulting insights.

  • Completion of HCGS Level 1 analysis. - June 1990
  • Completion of enhancements to the - Nov. 1990 SGS Level 1 analysis.
  • Completion of HCGS containment - Oct. 1991 performance analysis.
  • Completion of SGS containment - Oct. 1992 performance analysis.
  • Completion of final assessments - July 1993 and recommendations, submittal

- of the IPE report.

PSE&G will provide a brief summary report to the NRC upon completion of each milestone.

Page 4 of 4

Retrieved from "https://kanterella.com/w/index.php?title=ML18094B148&oldid=1940532"