ML20081K930
| ML20081K930 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 12/31/1994 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20081K932 | List: |
| References | |
| 25A5680, 25A5680-R01, 25A5680-R1, NUDOCS 9503290341 | |
| Download: ML20081K930 (62) | |
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TECHNICAL SUPPORT DOCUMENT FOR THE ABWR i
i General Electric Company SanJose, California December 1994 l
C503290341 941221 4
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25A5680 TECHNICAL SUPPORT DOCUMENT FOR THE ABWR TABLE OF CONTENTS Section Iidt Sheet No.
EXECUTIVE
SUMMARY
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1.0 INTRODUCTION
1.1 Background
6 1.2 Purpose 7
1.3 Description of Technical Support Document 8
2.0 EVALUATIONS OF RADIOLOGICAL RISK FROM NUCLEAR POWER PIANTS 2.1 Evaluation of SAMDAs under NEPA and Limerick Ecology Action 8
2.2 Cost / Benefit Standard for NEPA Evaluation of SAMDAs 9
2.3 Socio-Economic Risks for Severe Accidents 10 3.0 RADIOLOGICAL RISK FROM SEVERE ACCIDENTS IN PLANTS OF ABWR DESIGN 3.1 Severe Accidents in Plants of ABWR Design 11 3.2 Dominant Severe Accident Sequences for Plants of ABWR Design 13 3.3 Overall Conclusions from Chapter 19 of the ABWR SSAR 14 4.0 COST /BENEPT EVALUATION OF SAMDAS FOR PLANTS OF BWR DESIGN 4.1 SAMDA De2nition Applied to Plants of ABWR Design 14 4.2 Cost / Benefit Standard for Evaluation of ABWR SAMDAs 14 4.3 Candidate SAMDAs for the ABWR Design 15 4.4 Cost Estimates of Potential Modifications to the ABWR Design 15 4.5 Benefits of Potential Modifications to the ABWR Design 16 4.6 Cost / Benefit Comparison of SAMDAs 16 5.0
SUMMARY
AND CONCLUSIONS 16
6.0 REFERENCES
16 ATTACHMENTA Evaluation of Potential Modifications to the ABWR Design 31 2
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LIST OFTABLES Table life heet No.
1 Radiological Consequences of ABWR Accident Sequences 18
-2 Severe Accident Mitigation Design Alternatives (SAMDAs)
Considered for the ABWR Design 19 3
SAMDAs Evaluated under NEPA for the ABWR 22 4
Cost Estimates of SAMDAs Evaluated for the ABWR Under NEPA 25 5
Benefit Estimates of SAMDAs Evaluated for the ABWR Under NEPA 27 6
Comparison of Estimated Costs and Benefits of SAMDAs Evaluated for the ABWR under NEPA 29 A-1 Radiological Consequences of ABWR Accident Sequences 54 A2 Core Damage Frequency Contributors 55 A-3 Modifications Considered 56 A-4 Modifications Evaluated 59 A-5 Summary of Benefits 60 61 A-6 Summary of Costs A-7 Summary of Results 62 3
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25A5680 EXECUTIVE
SUMMARY
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The term " severe accident" refers to those events which are "beyond the substantial coverage of i
- design basis events" and includes those for.which there is substantial damage to the reactor core
- whether or not there are serious off-site consequences. See Severe Accident Policy Statement,50 Fed. Reg. 32,138 and 32,139 (August 8,1985).
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For new reactor designs, such as the ABWR, the Nuclear Regulatory Commission (NRC),in, _
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satisfaction ofits severe accident safety requirements and guidance, is requiring, among other.
j things, the evaluation of design alternatives to reduce the radiological risk from a severe accident by preventing substantial core damage (i.e., preventing a severe accident) or by limiting releases from the containment in the event that substantial core damage occurs (i.e., mitigating the impacts of a severe accident).
The National Environmental Policy Act (NEPA) requires the consideration of reasonable alternatives to proposed major Federal actions significantly affecting the quality of the human environment, including alternatives to mitigate the impacts of the proposed action. In 1989, a Federal Court of Appeals determined that NEPA required consideration of certain design i
alternatives; namely, severe accident mitigation design alternatives (SAMDAs). See Limerick l
Ecolouv Action v. NRC. 869 F.2d 719 (3rd Cir.1989). The court indicated that "[SAMDAs) are,-
as the name suggests, possible plant design modifications that are intended not to prevent an i
accident, but to lessen the severity of the impact of an accident should one occur " Id. at 731.
The court rejected the use of a policy statement as an acceptable basis for closing out NEPA l
consideration of SAMDAs in a licensing proceeding, because, among other things, it was not a j
rule making. Id. at 739.
i Recently, the NRC Stafr expanded the concept of SAMDAs to encompass design alternatives to prevent severe accidents, as well as mitigate them. See NUREG-1437, " Generic Environmental l
Impact Statement for License Renewal of Nuclear Plants," (Volume I, p. 5-100). By doing so, the l
Staff makes the set of SAMDAs considered under NEPA the same as the set of alternatives to prevent or mitigate severe accidents considered in satisfaction of the Commission's severe i
accident requirements and pohcy.
This document provides the technical basis for determining the status of severe accident closure under NEPA for the ABWR design. The report concludes that there is an adequate technical basis for closure of severe accidents under NEPA for the ABWR design. The basis and conclusions are expected to be codified in the form of proposed amendments to 10 CFR Part 52.
l The amendments would provide that:
j (1) For the ABWR design, all reasonable steps have been taken to reduce the occurrence of a i
severe accident involving substantial damage to the core and to mitigate the consequences of such an accident should one occur; i
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.(2) No cost <ffective SAMDAs to the ABWR design have been identified to prevent or mitigate
~ he consequences of a severe accident involving substantial damage to the core, j
t (3) No further evaluation of severe accidents for the ABWR design, including SAMDAs to the design, is required in any environmental report, environmental assessment, environmental impact statement or other environmental analysis prepared in connection with issuance of a combined license for a nuclear power plant referencing a certified ABWR design; and, j
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1.0 INTRODUCTION
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1.1 Background
The term " severe accident" refers to those events that are "beyond the substantial coverage of ;
design basis events" and includes those for which there is substantial damage to the reactor core whether or not there are serious off-site consequences. See Severe Accident Policy Statement,50
.t Fed. Reg. 32,138 and 32,139 (August 8,1985). For new reactor designs, such as the ABWR, the
' Nuclear Regulatory Commission (NRC), in satisfaction ofits severe accident safety requirements, l
is requiring, among other things, the evaluation of design alternatives to reduce the radiological -
risk from a severe accident by preventing substantial core damage (i.e., preventing a severe accident) or bylimiting releases from the containment in the event that substantial core damage occurs (i.e., mitigating the impacts of a severe accident).
The Commission's severe accident safety requirements for new designs are set forth in 10 CFR Part 52, $52.47(a)(1)(ii), (iv) and (v). Paragraph 52.47(a)(1)(ii) references the Commission's Three Mile Island safety requirements in $50.34(f). Paragraph 52.47(a)(1)(iv) concerns the treatment of unresolved safety issues and generic safety issues. Paragraph 52.47(a)(1)(v) requires the performance of a design-specific probabilistic risk assessment (PRA). The Commission's Severe Accident Policy Statement elaborates what the Commission is requiring for new designs.
The Commission's Safety Goal Policy Statement (51 Fed. Reg. 30,028 (August 21,1986)) sets goals and objectives for determining an acceptable level of radiological risk.
As part ofits application for certification of the ABWR design, GE has prepared a Standard Safety Analysis Report (ABWR SSAR). Chapter 19 of the ABWR SSAR, " Response to Severe Accident Policy Statement," demonstrates how the ABWR design meets the Commission's severe accident safety requirements and policies. In particular, Chapter 19 includes:
l (1) Identification of the dominant severe accident sequences and associated source terms for the ABWR design; (2) Descriptions of modifications that have been made to the ABWR design, based on the results of the Probabilistic Risk Assessment (PRA), to prevent or mitigate severe accidents and reduce the risk of a severe accident; (3) Bases for concluding that "all reasonable steps [have been taken] to reduce the chances of occurrence of a severe accident involving substantial damage to the reactor core and to mitigate the consequences of such an accident should one occur," (Severe Accident Policy Statement (50 Fed. Reg. 32,139)); and (4) Bases for concluding that the ABWR meets Commission's Safety Goals and objectives as set forth in the Safety Goal Policy Statement r
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Consequently, the conclusions are drawn in Chapter 19 that further modifications to the ABWR design to reduce severe accident risk are not warranted. The National Environmental Policy Act (NEPA) requires the consideration of reasonable alternatives to proposed major Federal actions significantly affecting the quality of the human emironment, including alternatives to mitigate the impacts of the proposed action. In 1989, a Federal Court of Appeals determined that NEPA required consideration of certain design alternatives; namely, severe accident mitigation design alternatives (SAMDAs). Limerick Ecolocv Action v. NRC,869 F.2d 719 (3rd Cir.1989). The court indicated that "[SAMDAs] are, as the name suggests, possible plant design modifications that are intended not to prevent an accident, but to lessen the severity of the impact of an accident should one occur." Id. at 731. The court rejected the use of a policy statement as an acceptable basis for closing out NEPA consideration of SAMDAs in a licensing proceeding, because, among other things,it was not a rule making, see id. at 739.
Subsequent to the Limerick decision, the NRC issued Supplemental Final Environmental Impact Statements for the Limerick and Comanche Peak facilities that considered whether there were any cost-effective SAMDAs that should be added to these facilities ("NEPA/SAMI)A FES Supplements"). On the basis of the evaluations in the supplements (called "NEPA/SAMDA evaluations"), the NRC determined that further modifications would not be cost-effective and were not necessary in order to satisfy the mandates of NEPA.
In recognition of the Limerick decision, the Commission is requiring NEPA consideration in Part 52 licensing of whether there are cost-effective SAMDAs that should be added to a new reactor design to reduce severe accident risk. While this consideration could be done later on a facility-specific basis for each combined license application under Subpart C to Part 52, the Commission has decided that maintenance of design standardization will be enhanced if this is done on a generic basis for each standard design in conjunction with design certification. See SECY-91-229,
" Severe Accident Mitigation Design Alternatives for Certified Standard Designs." That is, the Commission has decided to resolve the NEPA/SAMDA question through rule-making at the time of certification in a so called unitary proceeding, rather than in the context oflater licensing proceedings.
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Recently, the NRC Staff expanded the definition of SAMDAs to encompass design alternatives to prevent severe accidents, as well as mitigate them. See NUREG-1437," Generic Environmental Impact Statement for License Renewal of Nuclear Plants," (Volume I, p. 5-100). By doing so, the Staff makes the set of SAMDAs considered under NEPA the same as the set of alternatives to prevent or mitigate severe accidents considered in satisfaction of the Commission's severe accident requirements and policies.
1.2 Purpose The purpose of this technical support document is to provide a basis for determining the status of severe accident closure under NEPA for the ABWR design. The document supports a determination, which could be codified in a manner similar to the format of the Waste l
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' Confidence Rule (10 CFR $51.23), as proposed in amendments to 10 CFR Part 52. These amendments would provide that:
(1) For the ABWR design, all reasonable steps have been taken to reduce the occurrence of a -
severe accident involving substantial damage to the core and to mitigate the consequences of such an accident should one occur; (2) No costeffective SAMDAs to the ABWR design have been identified to prevent or mitigate the consequences of a tevere accident involving substantial damage to the core; (3) No further evaluation of severe accidents for the ABWR design, including SAMDAs to the design, is required in any environmental report, environmental assessment, environmental impact statement or other environmental analysis prepared in connection with issuance of a combined license for a nuclear power plant referencing a certified ABWR design; and, The evaluation presented in this document is modeled after that found in the Limerick and Comanche Peak NEPA/SAMDA FES Supplements for those facilities. Additionalinformation concerning the radiological risk from severe accidents for those plants is not found in the supplements, but in the FESS for the Limerick and Comanche Peak facilities. That information with respect to the ABWR design is presented in this document. The discussion herein of the radiological risk from severe accidents is based on Chapter 19 of the ABWR SSAR. Attachment A to this document presents the basis for concluding that further modifications to the ABWR design are not warranted in order to reduce the risk of a severe accident through the addition of design features to prevent or mitigate a severe accident. This information originally appeared as Appendix P to Chapter 19 of the SSAR. It was subsequently agreed with the NRC staff that this information should be set forth in an attachment to this document; accordingly, it has been located, in updated form, as Attachment A. hereto.
1.3 Description of Technical Support Document Section 2.0 provides an overview of the radiological risks from severe accidents. Sections 3.0 through 5.0 provide the NEPA/SAMDA analysis. Section 3.0 discusses the methodological approach to the evaluation ofSAMDAs under NEPA. Section 4.0 presents the results of the cost-effectiveness evaluation of the potential SAMDA modifications. Section 5.0 presents the conclusions and Section 6.0 the references.
2.0 EVALUATIONS OF RADIOLOGICAL RISK FROM NUCLEAR POWER PLANTS 2.1 Evaluation of SAMDAs Under NEPA and IAnerick Ecology Action Limerick Ecoloav Action stands for two propositions. First, NEPA requires explicit consideration of SAMDAs unless the Commission makes a finding that the severe accidents being mitigated are remote and speculative. Second, the Commission may not make this finding and dispose of 1
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NEPA consideration of SAMDAs by means of a policy statement. The purpose of evaluating SAMDAs under NEPA is to assure that all reasonable means have been considered to mitigate the impacts of severe accidents that are not remote and speculative. As discussed above, the Commission has indicated that it will resolve the NEPA/SAMDA issue for a new reactor design in the same proceeding, called a unitary proceeding, in which it certifies that design.
The Commission's Severe Accident and Safety Goal policy statements require the Commission to make certain findings about each new reactor design. For evolutionary designs, of which the ABWR is one, this must be done by the Staffin conjunction with FDA approval and by the Commission in conjunction with certification. First, the Commission must find that an evolutionary plant meets the safety goals and objectives; i.e., that the radiological risk from operating an evolutionary plant will be acceptable, meaning that any further reduction in risk will not be substantial.
Second, the Commission must find that all reasonable means have been taken to reduce severe accident risk in the evolutionary plant design. As part of the basis for making this finding, the cost-effectiveness of risk reduction alternatives of a preventive or mitigative nature must be evaluated.
Chapter 19 of the ABWR SSAR demonstrates that these findings can be made for the ABWR design. Given the nature and findings of these severe accident and safety goal evaluations, GE believes that a suflicient basis exists for finding by rule that further consideration of severe accidents, including evaluation of SAMDAs pursuant to NEPA, is neither necessary nor i
reasonable.
2.2 Cost / Benefit Standard for NEPA Evaluadon of SAMDAs The Limerick decision interpreted NEPA to require evaluation of SAMDAs for their risk j
reduction potential. In implementing the court's decision, the NRC considered the cost-effectiveness of each candidate SAMDA in mitigating the impact of a severe accident, using the i
$1,000 per person-rem averted standard. This standard is a surrogate for all oiT-site l
consequences.
The basic approach in this study is to rank the SAMDAs in terms of their cost-effectiveness in mitigating the impact of a severe accident. The criterion applied is the $1,000 per person-rem i
averted standard, which is what the Commission has historically used in distinguishing among and ranking design alternatives, including SAMDAs.
The Commission has used this standard in the context of both safety and NEPA analyses. For example, in the context of safety analysis, the standard has been used to perform evaluations associated with implementation of the Safety Goal Policy Statement; the Severe Accident Policy Statement; and $50.34(f) requirements. In the context of environmental analysis,it has been l
used in the Limerick and Comanche Peak NEPA/SAMDA FES Supplements; and in the draft Generic Environmental Impact Statement for License Renewal of Nuclear Plants (NUREG-1437).
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- f As indicated above, the Commission is preparing i Generic Environmental Impact Statement for l
License Renewal of Nuclear Plants.: The draft statement, NUREG 1437, makes clear that the use -
. of this standard in the evaluation of severe accident risk reduction alternatives, which include ;
SAMDAs, is acceptable (see NUREG-1437, Vol. I, p. 5-108).
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~- On the basis of these considerations, the cost / benefit ratio of $1,000 per person-rem averted is viewed as an acceptable standard for the purposes of evaluating SAMDAs under NEPA.~
2.3 Socio Economic Risks for Severe Accidents
=i As discussed above in Section 2.2, the Commission uses the $1,000/ person-rem-averted standard as a surrogate for all off-site consequences. See SECY-89-102, " Implementation of Safety Goal l;
Policy." However, Environmental Impact Statements (EIS) for nuclear power plants provide
-l separate, general discussions of the socio-economic risks from severe accidents. In keeping with this precedent, GE is providing a general discussion of socio<conomic risks for the ABWR design, i
based in large measure on the discussion of such risks in NUREG-1437, " Generic Environmental Impact Statement for License Renewal of Nuclear Plants."
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The term "socio-economic risk from a severe accident" means the probability of a severe accident 3
multiplied by the socio<conomic impacts of a severe accident. "Socio economic impacts,"in turn, relate to off-site costs. The off-site costs considered in NUREG-1437 (see W. I, p. 5-90) are:
Evacuation costs Value of crops or milk, contaminated and condemned Costs of decontaminating property where practical
-l Indirect costs due to the loss of the use of property or incomes derived therefrom (including
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interdiction to prevent human injmy), and Impacts in wider regional markets and on sources of supply outside the contaminated area.
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'l NUREG-1437 estimated the socioeconomic risks from severe accidents. The estimates were i
based on 27 FESS for nuclear power plants that contain analyses considering the probabilities and consequences of severe accidents. For these plants, the off-site costs were estimated to be as high as $6 billion to $8 billion dollars for severe accidents with a probability of once in one million operating years of occurring. Hig ier costs were estimated for severe accidents with much lower probabilities. The projected cost of adverse health effects from deaths and illnesses were estimated to average about 1020% of off-she mitigation costs and were not included in the $6-$8 billion dollar estimate.
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4 Another source of costs, which NUREG-1437 indicated could reach into the billions of dollars, was costs associated with the termination of economic activities in a contaminated area, which would create adverse economic impacts in wider regional markets and sources of supplies outside the contaminated area. The predicted conditionalland contamination was esumated to be small.
(10 acres / year at most). (See NUREG-1437, Vol. I, pp. 5-90 through 5-93.)
NUREG-1437 provides the bases for concluding that the socio-economic risks from severe accidents are predicted to be small and the residual impacts of severe accidents so minor that i
detailed consideration of mitigation alternatives is not warranted. See 56 Fed. Reg. 47,016,.
i 47,019,47,034 and 47,035 (September 17,1991).
The socic> economic risks contained in NUREG-1437 are bounding for plants of ABWR design.
First. the core damage frequency for plants of ABWR design is 1.6E-7 per year. Thus, no accidents, and hence no off-site costs, are expected at probabilities at or greater than once in one million years. Second, plants of ABWR design meet the safety goals set forth by the NRC. See Section 3.2, below.
3.0 RADIOLOGICAL RISK FROM SEVERE ACCIDENTS IN PLANTS OF ABWR DESIGN 3.1 Severe Accidents in Plants of ABWR Design Chapter 19 of the ABWR SSAR," Response to Severe Accident Policy Statement," establishes that the Cornmission's severe accident safety requirements have been met for the ABWR design, including treatment ofinternal and external events, uncertainties, performance of sensitivity studies, and support of conclusions by appropriate deterministic analyses and the evaluations,
required by 10 CFR Part 50.34(0, It also establishes that the Commission's safety goals have been
- met, l
Specifically, the following topics were addressed in Chapter 19 of the ABWR SSAR:
I (1) Consideration of the contributions ofinternal events (Section 19.3), Shutdown events j
(Section 19.4) and external events (Section 19.4) to severe accident risks, including a seismic risk analysis based on the application of the seismic margins methodology (Appendix 191);
(2) Identification of the ABWR dominant accident sequences; (3) Identification of severe accident risk reducdon features which were included in the ABWR design to achieve accident prevention and mitigation (addressed in Subsection 19.7.3(2));
Consideration of additional modifications, evaluated in accordance with $50.34(0(1), is addressed in Attachment A. Chapter 19 concludes that the severe accident requirements of10 CFR Part 52 ($52.47 (a)(1)(ii), (iv) & (v)) and the Severe Accident Policy Statement have been 11 Rev1
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j 25A5680 met.' It also provides a summary of the bases for these conclusions. In particular, Chapter 19 presents a summary of the bases for concluding that the requirements of 8 50.34(f) (referenced
. in $52.47(a)(1)(ii)) have been met, including $50.34(f)(1)(i), which requires " perform [ance of]
a plant / site 4pecific [PRA), the aim of which is to seek such improvements in the reliability of
- core and containment heat removal systems as are significant and practical and do not impact excessively on the plant." Attachment A presents 'the bases for concluding that further modifications to the ABWR design are not warranted in order to reduce the risk of a severe accident through the addition of design features to prevent or mitigate a severe accident.
Section 19.6 of the ABWR SSAR addresses how the goals of the Severe Accident Policy Statement have been met for plants of ABWR design. These goals include:
l Prevention of core damage
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Prevention of early containment failure for dominant accident sequences j
Evaluation of the effects of hydrogen generation U
Heat removal to reduce the probability of containment failure Prevention of hydrogen deflagration and detonation Offsite dose, and Containment conditional failure probability.
Specific conclusions concerning severe accidents for plants of ABWR design based on the ABWR SSAR Chapter 19 cvaluations are as follows:
(1) Catc, Damage Freauencv. The ABWR core damage frequency was determined to be 1.6E-7 per t eactor year in Subsection 19.6.2. The goal was IE 6 per reactor year.
(2) Conditional Containment Failure Probabilitv. The conditional containment failure probability was shown to be 0.002 in Subsection 19.6.8. This is significantly below the goal of 0.1.
(3) Individm1 Risk (Promot Fatality rim. The prompt fatality risk to a biologically average individual within one mile of an ABWR site boundary was determined to be 1.4E-13 per individual per year in Section 19E.3. This is significantly less than the goal of one tenth of one percent of the sum of prompt fatality risks resulting from other accidents to which members of the U.S. Population are generally exposed. The numerical value of this goalis 3.9E-7 per individual per year (or 0.04 per 100,000 people per year).
(4) Societal Risk (f ntent Far21ity Risk). The latent fatality risk to the population within 50 miles of an ABWR site boundary was determined to be 9.0E-13 per individual per year in Section 19E.3. This is significantly less than the goal of one tenth of one percent of the sum of the cancer fatality risks resulting from all other causes. The numerical value of this goal is 1.7FA per individual per year (or 0.17 deaths per 100,000 people per year).
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25A5680 (5) Probability of Larce Ofr-Site Dose. The probability of exceeding a whole body dose of 25 rem at a distance of one-half mile from a ABWR was determined to be less than IE-9 per reactor year in Section 19E.3.
Residual radiological risk from severe accidents in plants of ABWR design is summarized in Table A-1 (reproduced here as Table 1). The cumulative exposure risk to the population within 50 miles of a plant of ABWR design is approximately 0.269 person-rem for an assumed plant life of 60 years. This calculation includes the dominant sequences, as well as several sequences that are considered remote and speculative.
3.2 Dominant Severe Accident Sequences for Plants of ABWR Design in performing the PRA for the ABWR design, GE identified and evaluated many severe accident sequences. For each sequence, the analysis identified an initiating event and traced the accident's progression to its end. For sequences involving core damage, conditional containment failure probabilities and offsite consequences were estimated. After the accident scenarios were binned according to radiological release (source term) parameters, only two dominant cases remained.
The dominant cases are: Case 1 (best estimate core damage sequences that had rupture disk activation); and the NCL case (core damage with normal containment leakage). The residual risks of these two cases can be found in Table 1. The complete radiological consequence analysis of the dominant sequences can be found in Section 19E.3 of the ABWR SSAR.
The probability of occurrence of dominant sequences is greater than IE-9 per year. Several sequences with occurrence probabilities less than IE-9 per year were carried through the severe accident analysis in order to determine the sensitivity of plants of ABWR design to certain phenomena and parameters. These sequences were also considered in the SAMDA evaluation for sensitivity purposes.
Sequences with probabilities of occurrence less than IE-9 were considered remote and speculative. While the Commission has not yet specified a quantitative point at which it will consider severe accident probabilities as remote and speculative,it has indicated that a decision to consider severe accidents remote and speculative would be based upon the accident probabilities and the accident scenarios being analyzed. See Vermont Yankee Nuclear Power Corooration. (VermontYankee Nuclear Power Station), CLI-90-07,32 NRC 129,132 (1990).
GE believes that the severe accident analysis in Chapter 19 of the ABWR SSAR provides a suflicient basis for the Commission to find that ABWR sequences that are not dominant can be deemed remote and speculative.
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25A5680 3.3 Overall Conclusions from Chapter 19 of the ABWR SSAR The specific conclusions about severe accident risk discussed above support the overall conclusion that the environmental impacts of severe accidents for plants of ABWR design represent a low risk to the population and to the environment. For the ABWR design, all reasonable steps have been taken to reduce the occurrence of a severe accident involving substantial damage to the core and to mitigate the consequences of such an accident should one occur. No further cost-efTective modifications to the ABWR design have been identified to reduce the risk from a severe accident involving substantial damage to the core. No further evaluation of severe accidents for the ABWR design is required to demonstrate compliance with the Commission's severe accident requirements or policy or the safety goal.
4.0 COST / BENEFIT EVALUATION OF SAMDAS FOR PLANTS OF ABWR DESIGN 4.1 SAMDA Definition Applied to Plants of ABWR Design Attachment A considers whether the ABWR design should be modified in order to prevent or mitigate the consequences of a severe accident in satisfaction of the NRC's severe accident requirements in 10 CFR Parts 50 & 52 and the Severe Accident Policy Statement. The cost / benefit evaluation of SAMDAs to plants of ABWR design uses the expanded definition of SAMDAs set forth in NUREG-1437: design alternatives that could prevent and/or mitigate the consequences of a severe accident.
4.2 Cost / Benefit Standard for Evaluation of ABWR SAMDAs As discussed in Section 2.2 above, the cost / benefit ratio of $1,000 per person-rem averted is viewed by the NRC and the nuclear industry as an acceptable standard for the purposes of evaluating SAMDAs under NEPA. This standard was used as a surrogate for all off-site costs in the cost / benefit evaluation of SAMDAs to plants of ABWR design. Averted on-site costs were incorporated for SAMDAs that were at least partially preventive in naturel. On-site costs resulting from a severe accident include replacement power, on-site cleanup costs, and economic loss of the facility. A more detailed discussion of averted on-site costs can be found in Attachment A.
The equation used to determine the cost / benefit ratio is:
Cost of SAMDA implementation MINUS averted on-site costs Cost / benefit ratio
=
Reduction in residual risk (person-rem / plant life)
A plant lifetime of 60 years was assumed to maximize the reduction in residual risk.
1 Assessment of averted on-site costs are provided for information only. It is GE's position that the NRC is not required to account for these costs.
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25A5680 4.3 Candidate SAMDAs for the ABWR Design The complete list of SAMDAs considered for plants of ABWR design is contained in Table 2. This list is also contained in Table A-3 of Attachment A. The SAMDAs are classified according to the following categories:
(1) Modification is applicable to the ABWR and already incorporated into the design. No further evaluation is needed.
(2) Modification is applicable to the ABWR but not incorporated into the design. These modifications were considered further in Attachment A and the results of the cost / benefit analysis will be presented in this document.
(3) Modification is not applicable to the ABWR design due to the basis provided.
(4) Modification is considered as part of another modification listed in the table.
Table 3 lists the advantages and disadvantages of each design alternative that is applicable to the ABWR but not incorporated into the design ("2" classification in Table 2). A detailed discussion of each alternative is contained in Section A.4 of Attachment A.
4.4 Cost Estimates of Potential Modifications to the ABWR Design Table 4 provides a brief explanation of the estimated costs of each design alternative applicable to the ABWR design. Details of the cost estimation methodology are provided in Section A.I.S.2 of Attachment A. As discussed in Attachment A, rough order of magnitude costs, biased in favor of making a modification, were assigned to each modification. The costs represent the incremental costs that would be incurred in a new plant rather than costs that would apply on a backfit basis.
The estimated costs of design alternatives that are, at least partially, preventive in nature were adjusted for averted on-site costs. This adjustment is included in the cost estimates in Table 4.
Design alternatives that are purely mitigative in nature are not assigned any averted on-site costs because these modifications do not significantly affect site clean up cost nor significantly lessen the plant investment loss. Section A.5 of Attachment A discusses the bases for assigning averted on-site costs in detail.
Considerable uncertainties prevent precise cost estimates because design details have not been developed and construction and licensing delays cannot be accurately evaluated. For purpose of this evaluation, all known or reasonably expected costs were accounted for in order that a reasonable assessment of the minimum cost could be obtained. Using a minimum cost favors implementation of a modification. Actual implementation costs are expected to be significantly higher than those used in this evaluation.
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4.5 Benefits of Potential Modifications to the ABWR Design j
l Table 5 summarizes the basis for assigning a benefit to each SAMDA. In general, benefits were j
estimated from the PRA results of Chapter 19 of the ABWR SSAR by considering which sequences are affected by each modification. Detailed discussion of the method for estimating benefit is =
provided in Section A.4 of Attachment A. The averted residual risk for each SAMDA is also given in Table 5.
4.6 Cost / Benefit Comparison of SAMDAs Table 6 summarizes the results of combining the cost estimates from Table 4 with the benefit estimates from Table 5. As is evident from Table 6, none of the SAMDAs requires further evaluation since the cost / benefit standard was not met. The closest design alternative exceeds the criteria by more than a factor of 1000.
On the basis of the small residual risk of a plant of ABWR design,0.269 person-rem for the entire plant life, a design modification would have to cost $269 or less in order to meet the standard of
$1,000 per person-rem averted.
5.0
SUMMARY
AND CONCLUSIONS A reasonable and comprehensive set of candidate SAMDAs relevant to the ABWR design was evaluated in terms of minimum costs, averted on-site costs and potential benefits. A screening criterion of $1,000 per person-rem averted was used to determine which alternatives, if any, were cost <ffective. None was found to meet the criterion. In fact, the implementation cost of a SAMDA would have to be less than $269 in order to pass. Given the low residual risk profile of the ABh% design, SAMDAs cannot be reasonably incorporated in a cost-effective manner.
On the basis of the foregoing analysis, further incorporation of SAMDAs into the ABWR design is not warranted. No further screening of SAMDAs is needed and no SAMDAs need be incorporated into ABWR design in satisfaction of NEPA.
6.0 REFERENCES
1.
ABWR Standard Safety Analysis Report,23A6100, Docket No.52-001, GE Nu&ar Energy.
2.
Assessment of Severe Accident Prevention and Mitigation Features, NUREG/CR-4920, Brookhaven National Laboratory, July 1988.
3.
Design and Feasibility of Accident Mitigatk,n Systems for Light Water Reactors, NUREG/CR-4025, R&D Associates, August 1985.
16 Rev1
y 25A5680
- 4. _ Evaluation of Proposed Modifications to the GESSAR II Design, NEDE 30640 (Proprietary),
. June 1984.
5.
Generic Environmental Impact Statement for License Renewal of Nuclear Plants, NUREG-1437, August 1991.
6.
" Issuance of Supplement to the Final Environmental Statement-Comanche Peak Steam Electric Station, Units 1 and 2", NUREG 0775 Supplement, December 15,1989.
7.
Severe Accident Risks: An Assessment for Five US Nuclear Power Plants, NUREG 1150, january 1991.
8.
" Supplement to the Final Environmental Statement-Limerick Generating Station, Units 1 and 2", NUREG 0974 Supplement, August 16,1989.
9.
Survey of the State of the Art in Mitigation Systems, NUREG/CR-3908, R&D Associates, December 1985.
- 10. Technical Guidance for Siting Criteria Development, NUREG/CR-2239, Sandia National.
Laboratories, December 1982.
- 11. Title 10, Code of Federal Regulations, Part 50 and 52.
- 12. 50FR32138, Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants, August,1985.
- 13. 50FR30028, Safety Goals for the Operations of Nuclear Power Plants; Policy Statement, August 1986.
F l
17 Rev1
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25A5680 Table 1 Radiological Consequences of ABWR Accident Sequences Whole Body Cumulative Exposure Probability Exposure,50 mile Risk Case
. (Event / Year)*
(Person-rem)
(Per-rem /60 Yr)
NCL 1.3E-07 9.60E3 0.075 1
2.1E-08 1.38E4 0.017 2
7.8E-11 8.33E3 0.00004 3
0 3.71E5 0.000 4
0 2.06E5 0.000 5
7.5E-12 9.34E4 0.00004 6
3.1E-12 2.42E6 0.004 7
3.9E-10 2.73E6 0.064 8
4.1E-10 3.20E6 0.079 9
1.7E-10 3.31E6 0.034 Total:
0.269
- Sequences with probabilities of occurrence less than IE-9 per year are considered remote and speculative.
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l' 25A5680 Table 2 Severe Accident Mitigation Design Alternatives (SAMDAs)* -
Considered for the ABWR Design Modincation Category 1.
ACCIDENT MANAGEMENT
- a. Severe Accident EPGs/AMGs 2
- b. Computer Aided Instrumentation 2
- c. Improved Maintenance Procedures / Manuals 2
- d. Preventive Maintenance Features 4
- c. Improved Accident ManagementInstrumentation 4
- f. Remote Shutdown Station 1
. g. SecuritySystem I
- h. Simulator Training for Severe Accident 4
2.
REACTOR DECAY HEAT REMOVAL
- a. Passive High Pressure System 2
- b. Improved Depressurization 2
- c. Suppression PoolJockey Pump 2
- d. Improved High Pressure Systems 1
- c. Additional Active High Pressure System 1
- f. Improved Low Pressure System (Firepump) 1
- g. Dedicated Suppression Pool Cooling I
- h. Safety Related Condensate Storage Tank 2
i.16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> Station Blackout Injection 4
- j. Improved Recirculation Model 4
3.
CONTAINMENT CAPABILFIY
- a. Larger Volume Containment 2
- b. Increased Containment Pressure Capacity 2
- c. Improved Vacuum Breakers 2
- d. Increased Temperature Margin for Seals 1
- e. Improved Leak Detection 1
- f. Suppression Pool Scrubbing 1
- g. Improved Bottom Penetration Design 2
- SAMDAs include both preventive and mitigative design alternatives 19 Rev1
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25A5680
.9 Table 2 (Condaued)
~
Modificadon
.I Categosy j
4.
CONTAINMENT HEAT REMOVAL
- a. Larger Volume Suppression Pool 2
i
- b. CUW Decay Heat Removal 1-I
- c. High Flow Suppression Pool Cooling 1
~d. Passive Overpressure Relief 1
1 5.
CONTAINMENT ATMOSPHERE MASS REMOVAL
.Y.
- a. High Flow Unfiltered Vent 3
- b. High Flow Filtered Vent 3
- c. Low FlowVent (Filtered) 2
'i
- d. Low Flow Vent (Unfiltered) 1 i
6.
COMBUSTIBLE GAS CONTROL
- a. Post Accident Inerting System 3
- b. Hydrogen Control byVenting 3
- c. Pre-inerting 1
- d. Ignition Systems 3
- c. Fire Suppression System Inerting 3
7.
CONTAINMENT SPRAY SYSTEMS -
- a. Drywell Head Flooding 2
- b. Containment Spray Augmentation 1
8.
PREVENTION CONCEPTS
- a. Additional Service Water Pump 2'
- b. Improved Operating Response I
- c. Diverse Injection System 4
- d. Operating Experience Feedback I
- c. Improved MSIV/SRV Design 1
9.
AC POWER SUPPLIES
- a. Steam Driven Turbine Generator 2
- b. Alternate Pump Power Source 2
- c. Deleted
- d. Additional Diesel Generator 1
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g, 25A5680 Table 2. (Condnued)
Modificadon Category 9.
(Continued)
- e. Increased Electrical Divisions 1-
- f. Improved Uninterruptable Power Supplies 1
- g. AC Bus Cross-ties I
- h. Gas Turbine-1 i
- i. Dedicated RHR (bunkered) Power Supply 4
l 10.
DC POWER SUPPLIES
- a. Dedicated DC Power Supply 2
- b. Additional Batteries / Divisions 4
- c. Fuel Cells 4
^
- d. DC Cross-ties 1
- e. Extended Station Blackout Provisions 1
11.
ATWS CAPABILTIY
- a. ATWS Sized Vent 2
- b. Improved ATWS Capability 1-12.
SEISMIC CAPABILITY
- a. Increased Seismic Margins 1
- b. Integral Basemat 3
13.
SYSTEM SIMPLIFICATION
- a. Reactor BuildingSprays 2
I
- b. System Simplification 1
- c. Reduction in Reactor Bldg Flooding 1
14.
CORE RETENTION DEVICES
- a. Flooded Rubble Bed 2
- b. Reactor Cavity Flooder 1
- c. Basaltic Cements 1
j l
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21 Revl l
o Jo 25A5680 Table 3 SAMDAs Evaluated Under NEPA for the ABWR Potential Improvement Advantages Disadvantages la. Severe Accident Improved arrest of core melt None EPGs/AMGs progress and prevention of containment failure.
Ib. Computer Aided Improved prevention of core Additional training Instrumentation melt sequences Ic. Improved Maintenance Improved prevention of core Increased documentation cost Procedures / Manuals melt sequences 2a. Passive High Pressure Improved prevention of core High cost of additional system System melt sequences 2b. Improved Improved utilization of Low Cost of additional equipment Depressurization Pressure systems for prevention of core melt sequences 2c. Suppression PoolJockey Improved prevention of core Cost of additional equipment Pump melt sequences 2d. Safety Related Availability following Seismic I)esign and structural costs Condensate Storage Tank events Sa. Larger Volume a.
Increases time before a.
High cost Containment (Double containment failure b.
Containment failure not Free Volume) b.
Increases time for prevented recovery c.
Minor radiological benefit since risks dominated by long lived isotopes 3b. Increased Containment a.
Eliminates large releases a.
Extreme costs Pressure Capability b.
High temperature failures (Sullicient pressure to not prevented withstand severe accidents)
Reduces probability of a.
Increased maintenance Sc. Improved Vacuum a.
Breakers (Redundant suppression pool bypass and equipment costs valves in each line)
Cost for equipment and 3d. Improved Bottom Head a.
Increased time for in-a.
Penetration Design vessel arrest analysis a.
High cost 4a. Larger Volume a.
Increases heat absorption Suppression Pool (Double capability within effective liquid volume) containment 22 Revi
4 m
y y-25A5680 Table 3 (Continued)
PotentialImprovement Advantages Disadvantages
'l 4a. (Continued) b.
Increases time for b.
Minor radiological recovery of systems benefit since risks c.
Increases time before dominated bylong lived containment failure isotopes Probability of drywell 5a. Low Flow Filtered Vent a.
Provides some scrubbing a.
of fission products if head failure is low head fails relative to the other b.
Reduces containment containment failure j
leakage if movable modes penetrations are l
degraded c.
low cost 7a. Drywell Head Flooding Improved prevention of Additional cost of (Firewater crosstie to core melt sequences equipment drywell head area) 8a Additional Service Water Improved prevention of -
Additional cost of Pump core melt sequences equipment -
9a. Steam Driven Turbine Improved prevention of Additional cost of Generator core melt sequences equipment 9b. Alternate Pump Power Improved prevention of Additional cost of Source core melt sequences equipment 10a. Dedicated DC Power Additional time before Marginal benefit Supply containment overpressure 1la. ATWS Sized Vent a.
Provides scrubbing of a.
Uncertain location I
fission products, except b.
Potential for inadvertent noble gases, which pass actuation through reactor building c.
Floods reactor building :
which greatly hinders site recovery after accident d.
Potential failure of electrical equipment in reactor building 13a. Reactor Building Sprays Reduced release of Uncertain location and (Firewater crosstie for fission products from unknown potential reactor building sprays)
Reactor Building consequences from inadvertent actuation 23 Revl
.p y.,
25A5680 Table 3 (Continued)
PotentialImprovement Advantages Disadvantages 14a. Flooded Rubble Bed Prevention of core-Small benefit over passive concrete interaction.
flooding system.
afTects t
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25A5680 Table 4 Cost Estimates of SAMDAs Evaluated for the ABWR Under NEPA Potential Estimated Improvement Cost Basis Minimum Cost la. Severe Accident Plant specific procedure preparation
$ 600,000 EPGs/AMGs beyond generic work by Owners' Group.
Ib. Computer Aided Software modifications and interface
$ 599,600 Instrumentation hardware. Credit for averted onsite cost included.
Ic. Improved Maintenance Procedure preparation. Credit for averted
$ 299,000 Procedures / Manuals onsite cost included.
2a. Passive High Pressure System hardware and installation
$ 1,744,000 System
($1,200,000), Building modification
($550,000). Credit for averted onsite cost included.
2b. Improved Depressurization Logic, pneumatic supplies, piping and
$ 598,600 qualification. Credit for averted onsite cost included.
2c. Suppression PoolJockey System hardware and electrical
$ 120,000 Pump connections. Credit for averted onsite cost included.
2d. Safety Related Condensate Structural analysis and material. Credit for
$ 1,000,000 Storage Tank averted onsite cost included.
Sa. LargerVolume Double current volume at $1200/ft.
$ 8,000,000 Containment (Double Free Analysis not included.
Volume) 1 4
Sb. Increased Containment Similar to Larger Volume Containment,
$ 12,000,000 l
Pressure Capability but denser rebar and labor required.
(Sufficient pressure to Assumed 50% higher cost withstand severe accidents)
Sc. Improved Vacuum Breakers Eight lines at $10,000 per line
$ 100,000 (Redundant valves in each line) l l
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Table 4 (Condnued)
Potendal Estl==ted Improvement Cost Basis IW=imian Cost 3d. Improved Bottom Head 205 drives at $1,000/ drive and $500,000 of.
$ 750,000 Penetration Design analysis 4a. LargerVolume Suppression Assumed to be the same as Larger Volume
$ 8,000,000 Pool (Double effective Containment liquid volume) 5a. Low Flow Filtered Vent Hardware and Testing program
$ 3,000,000 7a. Drywell Head Flooding Minor valve and piping modification with
$ 100,000 (Firewater crosstie to instrumentation drywell head area) 8a. Additional Service Water System hardware, power supplies and
$ 5,999,000 Pump support systems. Credit for averted onsite cost included.
9a. Steam Driven Turbine System hardware, cabling and structural
$ 5,994,300 f
Generator changes. Credit for averted onsite cost included.
9b. Alternate Pump Power 400 kW generator at $300/kW. Credit for
$ 1,194,000 Source averted onsite cost included.
i 10a. Dedicated DC Power 5000 ft' building structure addition at
$ 3,000,000 Supply
$500/ft' and cabling 11a. ATWS Sized Vent Instrumentation and cabling
$ 300,000 in addition to training 13a. Reactor Building Sprays Minor valve and piping modification with
$ 100,000 (Firewater crosstie for instrumentation.
reactor building sprays) 14a. Flooded Rubble Bed 1250 ft' of material at $1000/lb
$ 18,750,000 l
i l
26 Rev1
25A5680 Table 5 Benefit Estimates of SAMDAs*
Evaluated for the ABWR Under NEPA Potential Averted Risk Improvement Benefit Basis Person-REM la. Severe Accident 10% improvement in mitigative actions 0.015 EPGs/AMGs
~
1b. Computer Aided 10% improvement in preventative actions 0.01 Instrumentation Ic. Improved Maintenance 10% improvement in reliability of RCIC, 0.016 Procedures / Manuals HPCF, RHR and LPFL 2a. Passive High Pressure 90% reliable diverse additional high 0.069 System pressure system 2b. Improved Depressurization 50% reduction in manual depressurization 0.042 reliability 2c. Suppression PoolJockey 10% improvement in low pressure makeup 0.002 Pump reliability.
2d. Safety Related Condensate Arbitrary selection due to high suppression 0.01 Storage Tank pool availability.
Sa. LargerVolume Elimination of drywell head failure 0.15 i
Containment (Double Free sequences Volume)
Sb. Increased Containment Elimination of all cases except normal 0.16 Pressure Capability containmentleakage (NCL)
(Sufficient pressure to withstand severe accidents)
Sc. Improved Vacuum Breakers Elimination of Case 2 sequences 0.00004 (Redundant valves in each line) 3d. Improved Bottom Head 50% improvement in in-vessel arrest due to 0.057 Penetration Design additional available time 4a. Larger Volume Suppression Elimination of Case 9 sequences involdng 0.0002 Pool (Double effective loss of suppression pool cooling systems liquid volume)
- SAMDAs include both preventive and mitigative design alternatives i
l i
27 Revi I
1 25A5680 i
l Table 5 (Continued)
Potential Averted Risk Improvement Benefit Basis Person-REM 5a. Low Flow Filtered Vent Elimination of sequences invohing 0.014 initiation of containment rupture disc 7a. Drywell Head Flooding Reduction in high temperature 0.06 (Firewater crosstie to containment failure sequences and drywell drywell head area) head failure sequences 8a. Additional Service Water 10% improvement in reliability of RCIC, 0.016 Pump HPCF, RHR and LPFL due to improved support systems 9a. Steam Driven Turbine Improved effective availability of EDG 0.052 Generator 9b. Alternate Pump Power Similar to additional high pressure 0.069 Source for high pressure system. See 2a.
systems 10a. Dedicated DC Power Similar to additional high pressure 0.069 Supply system. See 2a.
11a. ATWS Sized Vent Reduction in Case 9 sequences 0.03 13a. Reactor Building Sprays 10% reduction in consequence of 0.017 (Firewater crosstie for sequences involving containment leakage reactor building sprays) 14a. Flooded Rubble Bed Elimination of sequences invohing core-0.001 concrete interaction.
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25A5680 4
Table 6 Comparison of Estimated Costs and Benefits on SAMDAs*
Evaluated for the ABWR Under NEPA Cost-Benefit Estimated Rado Minimum Cost Averted Risk
($K per Person-PotentialImprovement
($)
Person-rem rem) la. Severe Accident EPGs/AMGs
$ 600,000 -
0.015
$ 40,000 lb. Computer Aided Instrumentation
$ 599,600 0.01
$ 59,600 1c. Improved Maintenance
$ 299,000 0.016
$ 18,700 Procedures / Manuals 2a. Passive High Pressure System
$ 1,744,000 0.069
$ 25,270 2b. Improved Depressurization
$ 598,600 0.042
$ 14,250 2c. Suppression PoolJockey Pump
$ 119,800 0.002
$ 59,900 2d. Safety Related Condensate Storage
$ 1,000,000 0.01
$ 100,000 Tank Sa. Larger Volume Containment
$ 8,000,000 0.15
$ 53,300 (Double Free Volume)
Sb. Increased Containment Pressure
$ 12,000,000 0.16
$ 75,000 Capability (Sufficient pressure to withstand severe accidents)
Sc. Improved Vacuum Breakers
$ 100,000 0.00004
$ 2,500,000 (Redundant valves in each line) 3d. Improved Bottom Head
$ 750,000 0.057
$ 13,160 Penetration Design 4a. Larger Volume Suppression Pool
$ 8,000,000 0.0002
$ 40,000,000 (Double effective liquid volume) 5a. Low Flow Filtered Vent
$ 3,000,000 0.014
$ 214,300 7a. Drywell Head Flooding (Firewater
$ 100,000 0.06
$ 1,700 crosstic to drywell head area)
- SAMDAs include both preventive and mitigative design alternatives 29 Rev1
y 25A5680 Table 6 (Continued)
Cost-Benefit Estimated Ratio Minimum Cost Averted Risk
($K per Perso:>
r PotentialImprovement
($)
Person-rem rem) 8a. Additional Service Water Pump
$ 5,999,000 0.016
$ 375,000 9a. Steam Driven Turbine Generator
$ 5,994,300 0.052
$ 115,300 9b. Alternate Pump Power Source
$ 1,194,000 0.069
$ 17,300 10a. Dedicated DC Power Supply
$ 3,000,000 0.069
$ 43,500 lla. ATWS Sized Vent
$ 300,000 0.03
$ 10,000 1Sa Reactor Building Sprays
$ 100,000 0.017
$ 5,900 (Firewater crosstie for reactor i
building sprays) 14a. Flooded Rubble Bed
$ 18,750,000 0.001
$ 18,750,000 30 Rev1
25A5680 l
)
A*ITACHMENT A*
Evaluation of Potential Modifications to the ABWR Design A.1 INTRODUCTION AND
SUMMARY
This attachment provides a description of an evaluation of potential changes to the ABWR design in order to determine whether further modifications can bejustified.
A.I.1 Background o
The U.S. Nuclear Regulatory Commission's policy related to severe accidents requires,in part, that an application for a design approval comply with the requirements of10CFR50.34(f). Item (f)(1)(i) requires performance of a plant site-specific [PRA] the aim of which is to seek improvements in the reliability of core and containment heat removal systems as are significant and practical and do not impact excessively on the plant. Chapter 19 of the ABWR SSAR provides the base PRA of the ABWR plant.
To address this requirement, a review of potential modifications to the ABWR design, beyond those included in the Probabilistic Risk Assessment (PRA), was conducted to evaluate whether potential severe accident design features could bejustified on the basis of cost per person-rem avertcd.
This attachment summarizes the results of GE's review and evaluation of the ABWR design.
Improvements have been reviewed against conservative estimates of risk reduction based on the PRA and minimum order of magnitude costs, to determine what modifications are potentially attractive.
A.I.2 Evaluation Criteria The benefit of a particular modification was defined to be its reduction in the risk to the general public.
Offsite factors evaluated were limited to health effects to the general public based on total exposure (in person-rem) to the population within 50 miles of the site. Five representative US regions were evaluated for selected individual ABWR sequences by the CRAC2 code. The regional results were then averaged to determine the exposures. Consistent with the standard used by the NRC to evaluate radiological impacts, health effect costs were evaluated based on a value of $1,000 per-ofTsite person-rem averted due to the design modification.
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25A5680 1
The offsite costs for other items such as relocation oflocal residents, elimination ofland use and decontamination of contaminated land were not considered. Reductions in the risk ofincurring onsite costs including economic losses, replacement power costs and direct accident costs are i
considered in this evaluation as credits against in the cost of the modification.
I Based on the PRA results (Section A.2),82% of the offsite risk results from very low probability
)
events which have high consequence. The maximumjustifiable cost of a modification was determined to be $269. Therefore, based on this methodology, no modifications arejustifiable.
However, a variety of modifications were reviewed to establish the relative attractiveness of potential changes.
1 1
A.I 3 Methodology The overall approach was to estimate the benefit of modifications in terms of dollar cost per total person-rem averted. Underestimated costs and overestimated benefits were assessed in order to favor modifications. Because of the uncertainties in the methodology and the desire to address severe accidents with sensible modifications, this basis isjudged to be acceptable for purposes of l
this study.
A.I.3.1 Selection of Modifications Potential modifications were identified from a variety of previous industry and NRC sponsored studies of preventative and mitigative features which address severe accidents. Based on this composite list of modifications considered on previous designs, potential modifications were selected for further review based on being (1) applicable to the ABWR design, and (2) not included in the reference PRA.
Additional detail on the selection of modifications is provided in Section A.S.
A.I.3.2 Costs Basis Rough order of magnitude costs were assigned for each modification based on the costs of systems and system improvements determined by GE. These costs represent the estimated incremental costs that would be incurred in a new plant rather than costs that would apply on a backfit hasis. Section A.5 defines the cost estimates for cach of the modifications.
Even for a new plant such as the ABWR, relatively large costs (several million dollars) can be i
expected for some modifications if they involve modifications of the building structures or i
arrangement. This is because the cost oflabor and material is often a function of the building area required. For other modifications which involve minor hardware addition, the cost is often 32 Revl
e 25A5680 dominated by the need for procedure and training additions which can amount to hundreds of thousands of dollars.
The costs estimates were intentionally biased on the low side, but all known or reasonably expected costs were accounted for in order that a reasonable assessment of the minimum cost would be obtained. Actual plant costs are expected to be higher than indicated in this evaluation.
All costs are referenced to 1991 U.S. dollars. For modifications which reduce the core damage i
frequency, the costs of modifications (Section A.5) were further reduced by an amount proportional to the reduction present worth of the risk of averted onsite costs. Onsite costs j
include replacement pewer costs, direct accident costs (including onsite cleanup) and the economic loss of the facility. Evaluation of this creditincluded the following considerations:
(1) Accidents were assumed to occur at any time during the 60 year life of the plant. All onsite costs associated with the accident were evaluated as to their value at the time of the accident.
The economic risk of such onsite costs was evaluated as a function of time based on the onsite costs and the core damage frequency determined by the PRA. The plant core damage frequency was considered to be constant over the life of the plant. The economic risks were then evaluated based on the present worth of the time dependent economic risks.
(2) Replacement power was based on a rate of $.013/kW-h differential as har cost. The differential rate was assumed to be constant over the remaining life of the plant.
(3) The economic value of the facility at the time of the accident was based on a straight line depreciated value. The initial invested cost was taken at $1.4 Billion based on DOE cost guidelines.
(4) Accident costs for onsite cleanup and facility were evaluated based on escalated costs to the time of the accident. Reference accident costs to the facility were assumed to be $2 Billion.
(5) The economic evaluations were based on a discount rate of 8% and escalation factor of 35 A.1.3.3 Benefit Basis The cumulative risk of accidents occurring during the life of the plant was used as a basis for estimating the maximum benefit that could be derived from modifications. A particular modification's benefit was based on its effect on the frequency of events or associated offsite dose summarized in Tables A-1 and Table A-2. Dominant contributing failure probabilities were identified based on the PRA. Changes in these probabilities were estimated to evaluate the benefit of modifications. This basis is consistent with the approach taken in previous NRC evaluations. The cumulative offsite risk was evaluated over a 60 year plant life with no escalation in the evaluation criteria of $1,000/ person-rem.
Section A.4 summarizes each concept and estimated benefit for each individual potential modification. For each modification the cost per person-rem averted was evaluated to obtain the results of the individual evaluations. These conclusions are provided in Section A.7.
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A.1.4 Summary of Results Potentially attractive modifications were selected based on previous evaluations of potential prevention and mitigation concepts applicable during severe accidents. Of the modifications applicable to the ABWR design and which were not already implemented, twenty one were selected for additional review.
j i
None of the modifications considered met the $1,000/ person-rem averted criteria. The low evaluated frequency of core damage and subsequent release of radioactive material does not support modification to the ABWR based on costs in relationship to the benefit of averted j
exposures.
Since the most beneficial modification was evaluated to be several orders of magnitude higher than the criteria,it was concluded that no additional modifications are warranted in the ABWR design to address severe accidents. Furthermore, due to its magnitude it can be calculated that this conclusion will not be sensitive to variations in the assumptions used in the PRA results.
j A.2 SEVERE ACCIDENT RISK OF ABWR The reference design for this study was the ABWR PRA as presented in the internal events PRA (Section 19.3 of the ABWR SSAR). This evaluation accounts for features which were included in the current ABWR design-specifically to address severe accidents. These features and the reference description include:
Design Feature SSAR References (1) Firewater pump crosstie 5.4.7.1.1.10 (2) Passive containment flooder 9.5.12 (3) Gas turbine generator 9.5.11 (4) Overpressure Protection 6.2.5.2.6 A summary of the core damage frequency and offsite exposure frequency with these features included is shown in Table A-1. Event frequencies used in this evaluation were the same as assumed in the base PRA. The offsite exposures shown in Table A-1 were calculated by the CRAC2 code for release cases with similar consequences. The cases can be characterized as follows:
Case 1 Core Melt arrested in vessel or in Containment with actuation of containment rupture disk.
Case 2 Low Pressure Core Melt with suppression pool bypass and actuation of containment i
rupture disk.
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-1
! Case 3 High Pressure Core Melt with drywell Head failure and fire water spray initiation -
1
- Case 4 -
Suppression Pool Decontamination reduction (Not used).'
Case 5 Large Break LOCA without recovery and with actuation of containment rupture :
~ disk.
Case 6 High Pressure Core Melt with Drywell Head failure and no firewater spray initiadon.
Case 7 Low Pressure Core Melt with Drywell Head failure and no mitigation I
Case 8 High Pressure Core Melt with Early Containment failure.
Case 9 ATWS event with Drywell Head failure.
q NCL Normal Containment Leakage to Reactor Building.
The offsitt nposures for each case shown in Table A-1 were calculated by the CRAC2 code for five representative US regions for the selected individual ABWR sequences as discussed in Section 19E.3 of the ABWR SSAR.
Table A-2 provides additional detail on the individual contributors to the total core damage frequency. As indicated on Table A-2, the core damage frequency is dominated bylow pressure transient events (LCLP) (61.4%), followed by high pressure transient events (LCHP) (28.1%),
and station blackout sequences (SBRC) (10.3%).
Review of Table A-1 also indicates that the dominant contributors to the ABWR offsite exposure risk are the relatively low probability (less than 4E-10/yr), high consequence events (Cases 6 through 9) which contribute about 82% of the offsite exposure risk.
A.3 POTENTIAL ABWR MODIFICATIONS Potential modifications to the ABWR design were derived from a survey of various studies indicated in References A-1 through A-7 and the ABWR design process discussed in Section 19.7 of the ABWR SSAR. From these, a composite list of modifications was established. This list of potential modifications was reviewed to identify concepts which were already included in the ABWR design or which are not applicable.
Table A-3 summarizes the complete list of modifications and their classification according to the following categories:
35 Rev1
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y 25A5680 (1) Modification is applicable to ABWR and alreadyincorporated in the ABWR design. No further evaluation is needed.
(2) Modification is applicable to ABWR and not incorporated in ABWR design. (Table A-4 lists the Category 2 modifications which are evaluated further in this attachment.)
(3) Modification is not applicable to the ABWR design due to the basis provided.
(4) Modification is applicable to ABWR and is incorporated with the referenced modification.
A.4 RISK REDUCTION OF POTENTIAL MODIFICATIONS This section provides evaluations of the benefit: of potential modifications to the ABWR design identified in Table A-4. For each modification the basis for the evaluation and the concept is described. Table A-5 summarizes the benefit in terms of person-rem averted risk for each of the evaluated modifications.
A.4.1 Accident Management Accident management is a current topic under generic development within the Industry through the development of Accident Management Guidelines (AMGs) and revisions to Emergency Procedure Guidelines (EPGs). The following modifications are based on implementation of such generic activity.
A.4.t.1 Severe Accident EPGs/AMGs The symptom based EPGs, were developed by the BWR Owners Group following the accident at Three Mile Island, Unit 2. Currently the EPGs are under revision and accident management guidelines (AMGs) are being developed for severe accidents. These should provide a significant improvement which reduces the likelihood of a severe accident. Elements of these guidelines i
(such as containment pressure and temperature control guidelines) also deal with mitigating the effects of accidents.
In the ABWR PRA, Emergency Operating Procedures (EOPs) are based on these guidelines.
Additional extensions of the EPGs and EOPs could be made to address arrest of a core melt, emergency planning, radiological release assessment and other areas related to severe accidents.
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!Since the existing EPGs cover preventive actions and some mitigative actions, the incremental
- benefit of this item would be primarily mitigative. It wasjudged that the reliability of manual
= actions associated with mitigation could be improved by 10%, especially in use of core melt arrest j
processes. ; Failure rates for manually initiated mitigative systems were decreased by 10%, to
. estimate the benefit. The resulting offsite risk reduction is about 0.015 person-rem over 60 years.
A.4.1.2. Computer Aided Instrumentation Computer aided artificial intelligence can be added which provides attention to risk issues in man-machine interfaces. Significant computer assisted display and plant status monitoring is aircady part of the ABWR control room design. Additional artificial intelligence could be -
designed which would display procedural options for the operator to evaluate during severe accidents. The system would be an extension of ERIS to provide human engineered displays of the important variables in the EPGs and AMGs.
Operator actions are made significantly more reliable by new features such as Emergency Procedure Guidelines, Safety Plant Parameter Displays (SPDS), and training on simulators. If the improvements described in Subsection A.4.1.1 are assumed to be implemented, the incremental benefit of additional improvements is expected to be low. The reliability of manually initiated.
preventive systems was increased by 10% to estimate the benefit. The estimated increm' ental benefit over severe accident EPGs (Subsection A.4.1.1) is about 3% in core damage frequency l
(CDF) Because the improvement affects all release cases, the incremental benefit is about 0.01 person-rem.
A.4.1.3 Improved Maintenance Procedures / Manuals For the GE scope of supply this item would provide additional information on the components important to the risk of the plant. As a result ofimproved maintenance manuals and information it would be expected that increased reliability of the important equipment would occur. This item would be a preventative improvement which would address several system or components to l
difTerent degrees.
j Based on a 10% improvement in the reliability of the High Pressure Core Flooder (HPCF),
Reactor Core Isolation Cooling (RCIC), Residual Heat Removal (RHR) and Low Pressure Core Flooder (LPFL) systems, the CDF is reduced by about 9% which has a corresponding estimated l
person-rem reduction of about 0.016.
A.4.2 Decay Heat Removal Significant improvements in the reliability of ABWR high pressure systems have been made.
Among these are RCIC restart (NUREG 0737, II.K.S.13) and isolation reliability improvements (NUREG 0737, II.K.3.15). Additionally, the redundant HPCF is an improvement over early product lines which used the single HPCF system.
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-1 A.4.2.1 Passive High Pressure System This concept would provide additional high pressure capability to remove decay heat through a diverse isolation condenser type system. Such a system would have the advantage of removing not only decay heat, but containment heat if a similar system to that under consideration for the,
Simplified BWR (SBWR) is employed.
I The benefit of this system would be equivalent to an additional diverse RCIC system in addition to an additional containment heat removal system. The added system was assumed to be 90%
- reliable, designed to operate independent of offsite power and to be capable ofin-vessel core >
- melt arrest. Based on a reduction in the RCIC failure rate, the benefit is estimated at about 0.069 person-rem averted.
t A.4.2.2 Improved Depressurization This item would provide an improved depressurization system which would allow more reliable access to low pressure systems. Additional depressurization capability may be achieved through manually controlled, seismically protected, air powered operators which permit depressurization to be manually accomplished in the event ofloss of DC control power or control air events.
The ABWR high pressure core damage events represent about 28% of the total core damage frequency, but about 46% of the offsite exposure risk. The success of manual initiation was assumed to be improved by 50% and therefore the depressurization failure rate was reduced by a j
factor of 2. Based on this estimate of benefit offsite person-rem is reduced by about 23% and the -
estimated benefit is about 0.042 person-rem.
A.4.2.3 Suppression PoolJockey Pump This modification would provide'a small makeup pump to provide low pressure decay heat removal from the Reactor Pressure Vessel (RPV) using suppression pool water as a source. The return path to the suppression pool would be through existing piping such as shutdown cooling return lines.
The benefit of this modification would be similar to that provided by the firewater injection and spray capability, but it would have the advantage that long term containment inventory concerns would not occur.
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If the system could make low pressure coolant makeup systems 10% more reliable, significant reductions in CDF would not be achieved because other low pressure systems are already highly reliable. The estimated benefit is that CDF is reduced 2% and the averted risk would be 0.002 person-rem.
A.4.2.4 Safety-Related Condensate Storage Tank
,a The current ABWR design consists of a standard non-seismically qualified Condensate Storage Tank (CST). This modification would upgrade the structure of the CST such that it would be t
available to provide makeup to the reactor following a seismic event.
This modification only benefits the risks of core damage following seismic events. However, because the suppression pool provides an alternate suction source and the HCLPF for the suppression pool is relatively high (Appendix 191 of the ABWR SSAR), the dominant failure modes are not limited by water availability. Therefore the benefit of this modification is considered small. A benefit of 0.01 person-rem averted was arbitrarily chosen for an upgraded CST.
A.4.3 Containment Capability The ABWR containment is designed for about 45 psig internal pressure and includes a
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containment rupture disk which would relieve excessive pressure ifit develops during a severe accident. By providing the release point from the wetwell airspace, mitigation of releases are achieved through scrubbing of the fission products in the suppression pool.
t A.4.3.1 LargerVolume Containment This modification would provide a larger volume containment as a means to mitigate the effects of severe accidents. By increasing the size the containment could be able to absorb additional noncondensible gas generation and delay activation of the containment rupture disk or early j
containment failure, i
This item would mitigate the consequence of an accident by delaying the time before the evere accident source term is released and allowing more time for radioactive decay and recovery of systems. However, if recovery does not occur, eventual release is not prevented and if operation of the containment overpressure rupture disk does not occur, ultimately the containment will fait due to the long term pressurization caused by core concrete interaction and steam generation.
If sequences involving drywell head failure were eliminated (Cases 3,6,7,8 and 9), the offsite risks would be reduced by about 82% and about 0.15 person-rem would be averted.
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. A.4.3.2 Increased Containment Pressure Capacity
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The design pressure of the ABWR containmentis 45 psig. The containment rupture disk pressure and ultimate capability are significantly higher. By increasing the ultimate pressure.
t capability of the containment (including seals), the effects of a severe accident could be reduced -
or eliminated by delaying the time of release. If the strength exceeded the maximum pressure obtainable in a severe accident, only normal containment leakage would result.
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This modification would mitigate the event, not change the core damage frequency and the increased pressure capability may not be suflicient to contain the long term pressurization caused by core concrete interaction and steam ;,eneration. However, ifit were able to prevent all severe source term release except for normal containment leakage, the person-rem risk would be about 0.02 person eem/60 years. Therefore, the benefit would be about 0.16 person-rem.
A.4.3.3 Improved Vacman Breakers j
The ABWR design contains single vacuum breaker valves in each of eight drywell to wetwell vacuum breaker lines. The PRA included failure of vacuum breakers in Case 2 assuming operation of wetwell spray. This modification would reduce the probability of a stuck open-vacuum breaker by making the valves redundant in each line and climinate the need for operator action.
If Case 2 sequences were eliminated, the benefit of this modification would be about 0.00004 person-rem averted.
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A.4.3.4 Improved Bottom Head Penetration Design The ABWR design includes a 2-inch stainless steel drainline from the bottom of the RPV which is used to prevent thermal stratification in the RPV during operation and to provide cleanup of the i
bottom head by the CUW system. A carbon steel transition piece connects the drain line to the j
RPV. During a severe accident this transition piece may be susceptible to melting and may l
provide the earliest path for release of molten core material from the RPV to the containment.
The penetrations for the fine motion control rod drives in the ABWR also may provide a pathway for release from the RPV following a severe accident. Failure of the intemal blowout supports on
- the lower core plate, provided to eliminate the support structure in current generation BWRs, and welds of the drives at the bottom of the vessel may allow the CRDs to be partially ejected into the drywell during the severe accident which would provide a small pathway for release to the i
containment.
l The modification is to change the transition piece material to Inconel or Stainless Steel which has a higher melting point. By so doing, additional time would be available for recovery of core cooling systems. This modification also would establish external welds or restraints on the CRDs external to the vessel so that the drives would not be ejected following failure of the internal 40 Rev1
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benefit is not lost from eliminating the support beams in current generation BWRs. The benefit of these modifications would be to reduce the probability ofin-vessel arrest failure (NO IV).
Based.on consideration of the heatup rate of the bottom head,it has been estimated that making j
these changes could provide up to two hours additional time for recovery of systems. It is estimated, based on engineeringjudgment, that this time could result in the in-vessel arrest failure probabilities being reduced by a factor of two. The resulting benefit is about 0.057 person-rem averted.
~ A potential negative aspect of the modifications is that RPV failure could occur at another unknown location such as the bottom head itself. Although the time of vessel failure would be extended, the failure mode from these other locations could be potentially more energetic and lead to unevaluated consequences.
A.4.4 Containment Heat Removal i
The ABWR design contains 3 divisions of suppression pool cooling and provisions for a j
containment rupture disk for decay heat removal. In addition, modifications have been made to use the CUW heat exchangers to the maximum extent possible. Consequently, loss of containment heat removal events contribute only 0.1% of the total core damage frequency and l
offsite exposures. Additional modifications are not likely to show substantial safety benefits.
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A.4.4.1 LargerVolume Suppression Pool This item would increase the size of the suppression pool so that the heatup rate in the pool is reduced. The increased size would allow more time for recovery of a heat removal system.
Since this modification primarily affects LHRC events (Table A-2), the maximum benefitwould be climination of the LHRC contribution to the Case 9 sequences. These events are mitigated by the containment rupture disk and only contribute about 0.0002 person-rem to the base case risk.-
The assessed maximum benefit is therefore about 0.0002 person-rem.
A.4.5 Containment Atmosphere Mass Removal The ABWR design contains a containment rupture disk which provides containment overpressure protection from the wetwell airspace and utilizes the suppression pool scrubbing feature of the-suppression pool to reduce the amount of radioactive material released. One additional modification was considered.
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C 25A5680 A.4.5.1 Iow How Filtered Vent Some BWR facilities, especially in Europe, recently have added a filter system external to the containment to further reduce the magnitude of radioactive release. The systems typically use a multi-venturi scrubbing system to circulate the exhaust gas and remove particulate material. In the ABWR, because of the suppression pool scrubbing capability, a significant safety improvement
- is not expected due to this modification.
The release of radioactive isotopes from the ABWR following severe accidents occurs through the
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containment rupture disk for Cases 1,2 and 5. These sequences total about 8% of the exposure risk. The remaining sequences involve drywell head failure or early containment failure which would not be affected by this modification. The maximum benefit of the external vent system is therefore about 0.014 person-rem assuming perfect initiation of the filtered containment vent system.
A.4.6 Combustible Gas Control No additional modifications to the ABWR were identified in this group.
A.4.7 Containment Spray Systems A.4.7.1 Daywell Head Flooding This concept would provide intentional flooding of the upper drywell head such that if high drywell temperatures occurred, the drywell head seal would not fail. Additionally, if the seal were to fail due to overpressurization of the drywell, some scrubbing of the released fission products would occur. This system would be designed to operate passively or use an AC-independent water source.
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If an extension of the fire pump to drywell spray crosstie were considered for manual initiation of upper head flooding, additional reduction in the high temperature containment failure sequences (Case 8) would result. Additionally, a reduction in the high consequence drywell head failure sequences (Cases 6 and 7) could be achieved. If Case 8 sequences were eliminated and Case 6 and 7 source terms were reduced to a level similar to Case 3, the conservative benefit would be 0.12 penon-rem. The estimated benefit of this is about 0.06 person-rem assuming a 50% reliability ofinitiation.
A.4.8 Prevention Concepts The ABWR design contains an additional division of high pressure makeup capability to improve its capability to prevent severe accidents other features such as the fire pump injection capability and the combustion gas turbine have been included in the design to enhance the plant capability to prevent core damage. The following additional concepts were considered:
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s 25A5680 A.4.8.1 Additional Service Water Pumps This item addresses a reduction in the common cause dependencies through such items as improved manufacturer diversity, separation of equipment and support systems such as service water, air supplies, or heating and ventilation (HVAC). The HPCF, RCIC, and LPFL pumps are diverse in the ABWR design since they are either supplied by different manufacturers or have different flow characteristics. Equipment is separated in the ABWR design in accordance with Regulatory Guide 1.75. Thus, no further improvement is expected with regard to separation.
A reduction in common cause dependencies from support systems such as service wata stems, could conceivably reduce the plant risk through an improvementin system reliability. 'ihr concept for this item would be to provide an additional cooling water system capable of supporting each of the four divisional systems identified above.
The current design provides support to these systems from one of three divisions. Thus, the effect of this change would be to include a diverse and additional support system. In addition, diversity in instrumentation which controls these systems could be included so that redundant indication and trip channels would rely on diverse instrumentation.
A 10% increase in the reliability of the four systems was assumed which is the same improvement that may be derived from improved maintenance (Subsection A.4.1.3). This results in an estimated benefit of about 0.016 person-rem.
A.4.9 AC Power Supplies The current ABWR electrical design is improved through application of a gas-turbine generator to augment the offsite electrical grid. The following concepts were considered for additional onsite power supplies.
A.4.9.1 Steam Driven Turbine Generator A steam driven turbine generator could be installed which uses reactor steam and exhausts to the suppression pool. The system would be conceptually similar to the RCIC system with the generator connected to the offsite power grid.
The benefit of this item would be similar to the addition of another gas turbine generator, but would be somewhat less due to the relative unreliability of the steam turbine compared with a diesel generator and its unavailability after the RPV is depressurized. Ifit were sized large enough,it could have the advantage of providing power to additional equipment.
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A 25A5680 If the system has a 80% availability for all events, the benefit is similar to an 80% reduction in the diesel generator common mode failure rate. Evaluation of the PRAindicates that the resulting benefit is about 0.052 person-rem.
A.4.9.2 Alternate Pump Power Source The ABWR provides separate diesel driven power supplies to'the HPCF and LPFL pumps. Offsite power supplies the feedwater pumps. This modification would provide a small dedicated power source such as a dedicated diesel or gas turbine for the feedwater, or condensate pumps so that they do not rely on offsite power.
The benefit would be less dependence on low pressure systems during loss of offsite power events I
and station blackout events. If the feedwater system were made to be 90% available during loss of offsite power events and station blackouts, the benefit would be similar to adding an additional RCIC system (Subsection A.4.2.1). The resulting benefit would be about 0.069 person-rem.
A.4.10 DC Power Supplies
- The ABWR contains 4 DC divisions with sufficient capacity to sustain 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of station blackout (with some load shedding). This represents an improvement over current operating plant designs.
A.4.10.1 Dedicated DC Power Supply This item addresses the use of a diverse DC power system such as an additional battery or fuel cell '
for the purpose of providing motive power to certain components. Conceptually a fuel cell or.
separate battery could be used to power a DC motor / pump combination and provide high pressure RPV injection and containment cooling. With proper starting controls such a system could be sized to provide several days capability.
Providing a separate DC powered high pressure injecdon capability has a benefit of further reducing the station blackout and loss of offsite power event risks which represent about 75% of -
the total CDF, but only a small fraction of the offsite risk. If the effective unavailability of the RCIC is reduced by a factor of 10 due to the availability of a diverse system, one benefit would be similar to adding a power supply for feedwater (Subsection A.4.9.2) and the benefit would be about 0.069 person-rem.
t A.4.11 ATWS Capability The current ABWR design provides improvements in containment heat removal and detection of ATWS events to limit the impact of this class of events. The PRA indicates that ATWS events contribute about 0.1% of the core damage frequency (Table A-2) and about 17% of the offsite risk (Case 9).
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25A5680 A.4.11.1 ATWS Sized Vent This modification would be available to remove reactor heat from ATWS events in addition to severe accidents and Class 11 events. It would be similar to the containment rupture disk (which is currently sized to pass reactor power consistent with that generated during RCIC injection), but it
-would be of the larger size required to pass the additional steam associated with LPFL injection.
The system would need to be manually initiated.
The benefit of this venting concept is to prevent core damage and to reduce the source term available for release following ATWS events. The evaluation shows that an ATWS sized vent -
manually initiated with a 100% reliability would have a maximum benefit of reducing the offsite dose by about 0.03 person-rem by reassigning the consequences from Case 9 to Case 1.
A.4.12 Seismic Capability The current ABWR is designed for a Safe Shutdown Earthquake of 0.3g acceleration. The seismic margins analysis (Appendix 191 of the ABWR SSAR) addresses the margins associated with the seismic design and concludes that there is a 95% confidence that exis ting equipment has less than a 5% probability of failure at twice the SSE level. This capability is considered adequate for the ABWR design and no additional changes are considered.
A.4.13 System Simplification This item is intended to address system simplification by the elimination of unnecessary interlocks, automatic initiation of manual actions or redundancy as a means to reduce overall plant risk. Elimination of scismic and pipe whip restraints is included in the concept.
While there are several examples of redundant systems, valves and features on the ABWR design which could conceivably be simplified, there are several areas in which the ABWR design already has been improved and simplified, especially in the area of controls and logic. System interactions during accidents were included in this category. One area was identified in which simple modification of an existing system could provide some benefit.
A.4.13.1 Reactor Building Sprays This concept would use the firewater sprays in the reactor building to mitigate releases of fission products into the reactor building following an accident. The concept would require additional valves and nozzles, separate from the fire protection fusible links, to spray in areas vulnerable to release, such as near the containment overpressure reliefline routing.
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The benefit of this modification could be to reduce the impact of events which do not involve the operation of the containment rupture disk. Such events release fission products from the containment into the reactor building. Releases from normal containment leakage and cases 3, 6,7,8 and case 9 sequences could potentially be reduced. If 10% of these releases from these-cases were arbitrarily mitigated by this method, the benefit would be about 1.7E-04 person-rem.
A.4.14 Core Retention Devices i
Core retention features are incorporated into the ABWR Design. As discussed in Subsection
'19E.2.2(paragraph FS) of the ABWR SSAR, if a severe accident has resulted in a loss of RPV integrity, accident management guidance specifies that drywell sprays be initiated which will cause the suppression pool to overflow into the lower drywell after a few hours and quench the debris bed. After the molten core has been quenched, r3 further ablation of concrete is expected and the decay heat can be removed by normal.ontainment cooling methods such as suppression pool cooling. If sprays can not be initiated, the Lower Drywell Flooder System described in Subsection 9.5.12 of the ABWR SSAR cools a debris bed by flooding over the molten core in the lower drywell with water from the suppression pool. This system is similar to the Post Accident Flooding concept included in Reference A-4. One add,itional concept from Reference A 4 is included.
A.4.14.1 Mooded Rubble Bed This concept consists of a bed of refractory pebbles which fill the lower drywell cavity and are flooded with water. The bed impedes the flow of molten corium and increases the available heat transfer area which enhances debris coolability. The use of thoria (Th02) pellets in a multiple layer geometry has been shown to stop melt penetration; thus, preventing core-concrete j
interaction. Drawbacks to using thorium dioxide include cost, toxicity, and the radiological impact of radon gas release into the lower drywell via the radioactive decay of thorium. Other refractories such as alumina slow corium penetration but may fail to stop core-concrete contact.
Other refractories may be susceptible to chemical attack by the corium and may melt at lower temperatures. Pebbles composed of refractories other than thoria also may be susceptible to floating because they have lower density the Mc corium. A major drawback common to all need for further experimental testing in order to flooded rubble bed core retention systems,
s validate the concept in BWR applications.
The benefit of this modification lies in the potential elimination of core-concrete interaction and a corresponding decrease in non-condensable gas generation. Attachment 19EC to Appendix 19E cf the ABWR SSAR indicates a 90% certainty that debris on a concrete floor covered with wato al be co. "e.ble in the current ABWR design.
Only sequences in which no liquid injection to the drywell occurs will result in core-concrete interaction. A conservative estimate of the benefit of this concept over the existing design would be elimination of sequences with core-concrete interaction except those with containment 46 Revi
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A.5 COST IMPACTS OF POTENTIAL MODIFICATIONS t
As discussed in Subsection A.1.3.1, rough order of magnitude costs were assigned to each modification based on the costs of systems determined by GE. These costs represent the incremental costs that would be incurred in a new plant rather than costs that would apply on a backfit basis. Credit for the onsite costs averted by the modification are discussed in Subsection A.l.3.2. For each modification which reduces the core damage frequency an estimate of the impact was made and then applied to the potential averted offsite cost. This section summarizes the cost basis for each of the modification evaluated in Section A 4. This basis is generally the cost estimate less the credit for onsite averted costs. Table A-6 summarizes the results.
The costs were biased on the low side, but all known or reasonably expected costs were accounted for in order that a reasonable assessment of the minimum cost would be obtained. Actual plant costs are expected to be higher than indicated in this evaluation. All costs are referenced to 1991 U.S. dollars based on changes in the Consumer Price Index.
A.5.1 Accident Management A.5.1.1 Severe Accident EPGs/AMGs The cost of extending the EPGs would be largely a one-time cost which should be prorated over several plants if accomplished by the BWROG. Current industry activity is addressing this as part of Accident Management Guidelines (AMG). If plant specific, symptom based, severe accident emergency procedures were to be prepared based on AMGs, the cost would be atleast $600,000 for plant specific modifications to EOPs.
A.5.1.2 Computer-Aided Instrumentation Additional software and development costs associated with modifying existing Safety Plant Display Systems are estimated to cost at least $600,000 for a new plant. This estimate is based on assumed additions ofisolation devices to transmit data to the computer and in plant wiring. Because this modilication reduces the frequency of core damage events, a present worth of $400 onsite costs are averted and the cost basis is $599,600.
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e 25A5680 A.5.1.3 Improved Maintenance Procedures / Manuals The c ast of at least $300,000 would be required to identify components which should receive enhai ced maintenance attention and to prepare the additional detailed procedures or recom rnended information beyond that currently planned. Credit for reduction in onsite costs reduce s the cost basis to $299,000.
A.5.2 Decay Heat Removal A.5.2.1 Passive High Pressure System The cost of an additional high pressure system for core cooling would be extensive since itwould not only require additional system hardware which would cost at least $1,200,000, but it would also require additional building costs for space available for the system. Assuming the system could be located in the reactor building without increasing its height, building costs are estimated to be another $550,000. The credit for averted onsite costs is about $6,000 which brings the cost basis to $1,744,000.
A.5.2.2 Improved Depressurization The cost of the additional logic changes, pneumatic supplies, piping and qualification was estimated for the GESSAR II design (Reference A-1). A similar cost would be expected for the AllWR design. The coat is estimated to be at least $600,000 for an improved system for depressurization. This estimate assumes no building space increase for the added equipment.
The credit for averted onsite costs was evaluated to be $1,400 which makes the cost basis
$598,600.
A.5.2.3 Suppression PoolJockey Pump The cost of an additional small pump and associated piping is estimated at more than $60,000 including installation of the equipment. It is assumed that increases in power supply capacity and building space are not required. Controls and associated wiring could cost an additional $60,000 for a total cost of at least $120,000. A credit of $200 for averted onsite costs makes the cost basis
$119,800.
A.5.2.4 Safety Related Condensate Storage Tank Estimating the cost of upgrading the CST structure to withstand seismic events requires a detailed structural analysis and resultant material. It isjudged that the final cost increase would be in excess of $1,000,000. No credit for onsite cost averted was assumed for this modification.
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A.5.3 Containment Capability -
A.5.3.1 Iarger Volume Containment Doubling the containment volume requires an increase in the concrete and rebar. If structural costs of the containment can be made for $1,200/ft', doubling the containment volume without increasing its height, the cost would be at least $8,000,000. This estimate does notinclude reanalysis and other documentation costs. Since this modification is mitigative, no credit for onsite averted costs was assumed.
A.5.3.2 Increased Containment Pressure Capacity i
The cost of a stronger containment design would be similar in magnitude to increasing its size (Subsection A.5.3.1). If the costs are primarily due to denser rebar required during installation and additional analysis, an estimate of at least $12,000,000 could be required. Since this modification is mitigative, no credit for onsite averted costs was assumed.
A.5.3.3 Improved Vacuum Breakers The cost of redundant vacuum breakers including installation and hardware is estimated at more
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than $10,000 per line. Instrumentation associated with this modification is not included. For the eight lines the cost of this modification is more than $100,000. Since this modification is mitigative, no credit for onsite averted costs was assumed.
A.5.3.4 Improved Bottom Penetration Design The cost increase of using a stainless or inconel transition piece as opposed to carbon steel would be expected to be small in comparison to the engineering and documentation change costs associated with the change. Costs, associated with external welds and support for the CRDs is j
judged to be at least $1000 per drive. In addition, about $500,000 of analysis would be required j
to develop the changes. This would dominate the cost of this modification when applied to all 205 drives. Such changes are estimated to be at least $750,000.
Since this modification is mitigative, no credit for averted onsite costs applies.
A.5.4 Containment Heat Removal A.5.4.1 Larger Volume Suppression Pool This concept would result in similar costs as item Subsection A.5.3.1 for providing a larger containment. An estimate of $8,000,000 is assigned to this item.
A.5.5 Containment Atmosphere Mass Removal A.5.5.1 Low How Filtered Vent 49 Rev1
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- The cost of added equipment associated with the FILTRA system (excluding a test program) was estimated to be about $5,000,000 in Reference A-4. Although a detailed estimate was not prepared for the ABWR, an estimate of $3,000,000 has been assumed for the purpose of this evaluation.
Since this modification is mitigative, no credit for averted onsite costs applies.
3 A.5.6 Combustible Gas Control No additional modifications to the ABWR were identified in this group.
A.5.7 Containment Spray Systems A.5.7.1 Daywell Head Flooding An additional line to flood the drywell head using existing firewater piping would be a reladvely inexpensive addition to the current system. Instrumentation and controls to permit manual control from the control room would be needed. it is estimated that the total modification cost would be at least $100,000 for the engineering, piping, valves and cabling.
Because this modification is mitigative, no credit for averted onsite costs has been applied.
A.5.8 Prevention Concepts A.5.8.1 Additional Service Water Pump The use of diverse instrumentation would not presumably have a significant equipment cost, but there would be an increased cost of maintenance and spare parts due to less interchangeability and less standardization of procedures.
These costs, however, are probably low in comparison with the extra support systems for air supply and service water. Equipment, power supplies and structural changes to include these new systems are estimated to cost at least $6,000,000. A small credit for averted onsite costs makes the cost basis for this item $5,999,000, based on the benefits discussed in Subsections A.4.1.3 and A.5.1.3.
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t A.5.9.I ' Steam-Driven Tuaine Generator I
The cost of the system should be similar to that for the RCIC system, but additional cost would be needed for structural changes to the reactor building plus the generator and its controls. This item is expected to cost at least $6,000,000.
With credit for averted onsite costs, the cost basis for this item becomes $5,994,300.
A.5.9.2 Alternate Pump Power Source A typical feedwater pump for an ABWR sized plant could require a 4000 kWe sized generator, at
$300 per kWe, a separate diesel generator and the supporting auxiliaries could cost at least
$1,200,000. This cost would include wiring and installation of the alternate generator, but does not assume additional structural costs.
With credit for averted onsite costs, the cost basis for this item becomes $1,194,000.
A.5.10 DC Power Supplies A.5.10.1 Dedicated DC Power Supply Fuel cells are largely a developmental technology, at least in the large size range required for this application. In addition the process involves some risk of fire. To address these concerns a cost of at least $6,000,000 would be expected. A separate battery would be less expensive than fuel cells, but would involve additional space requirements which could make this modification more expensive than adding a diesel generator as discussed in Subsection A.5.9.2.
A battery bank capable of supplying.400 kwe would be about 50 times larger in capacity than the emergency batteries. This number of batteries would require at least 5,000 ft' of space, assuming extensive stacking and without concern for seismic response. At $500/ft' construction cost, the additional space required would amount to $2,500,000 for this modification. Additional costs would be required for DC pumps, cabling and instrumentation and controllers. A total cost would be at least $3,000,000.
51 Rev1
o 25A5680 A.5.11 ATWS Capability A.5.11.1 ATWS Sized Vent Larger piping and additional training would be required to extend the existing rupture disk feature to be available during an ATWS event. Additional instrumentation and cabling would be required to make the vent operable from the control room. It is estimated that the incremental-cost would be at least $300,000.
A.5.12 Seismic Capability No modifications were considered for this group.
A.5.13 System Simplification A.5.13.1 Reactor Building Sprays The cost of this modification isjudged to be similar to the concept of drywell head flooding (Subsection A.5.5.1) ifit only involves piping and valves which are tied into the firewater system.
An estimate of $100,000 has been assigned to this item.
Onsite cleanup costs also could be affected by this modification. If the cleanup costs were eliminated an averted cost would conservatively be about $5,000.
A.5.14 Core Retention Devices A.5.14.1 Flooded Rubble Bed Reference A-4 estimated that the refractory material needed for this modification would cost approximately $1,000/lb. If the lower drywell were filled with about 1.5 ft of this material, which would remain well below the service platform, at least 1250 ft' of material would be required. Ifit weighs 15 lb/ft', the material cost alone would amount to $18,750,000.
A.6 EVALUATION OF POTENTIAL MODIFICATIONS A ranking of the modifications by $/ person-rem averted is shown in Table A-7 based on the results and estimates provided in Sections A.4 and A.5.
The lowest cost / person-rem averted modification is more than 1600 times the target criteria of
$1,000 per person-rem averted. Clearly none of the modifications isjustifiable on the basis of costs for person-rem averted. This can be attributed to the low probability of core damage in the ABWR with the modifications to reduce risk already installed.
52 Revl
- W 9"
.25A5680 A.7 :
SUMMARY
OF CONCLUSIONS Potentially attractive modifications were identified from previous evaluations of potential -
_ prevention and mitigation concepts applicable during severe accidents and discussion with the NRC staff. Potential modifications were reviewed to select those which are applicable to the ABWR design and which have not already been implemented in the design. Of these modifications, twenty one were selected for additional review.
The low level of risk in the ABWR is demonstrated by the total 60 year offsite exposure risk of :
0.269 person-rem. At this level only modifications which cost less than $269 can bejustified.
i Based on this low level no modifications arejustified for the ABWR. Based on the PRA results, none of the modifications provided a substantial improvement in plant safety.-
A.8 REFERENCES A-1 Evaluation of Proposed Modifications to the GESSAR II Design, NEDE 30640 (Proprietary), June 1984.
A-2 Supplement to the Final Environmental Statement-Limerick Generating Station, Units 1 and 2, NUREG-0974 Supplement, August 16,1989 A-3 Issuance of Supplement to the Final Environmental Statement-Comanche Peak Steam Electric Station, Units 1 and 2, NUREG 0775 Supplement, December 15,1989 A-4 Smvey of the State of the Art in Mitigation Systems, NUREG/CR-3908, R&D Associates, December 1985 A-5 Assessment of Severe Accident Prevention and Mitigation Features, NUREG/CR-4920, Brookhaven National Laboratory, July 1988.
A-6 Design and Feasibility of Accident Mitigation Systems for Light Water Reactors, i
NUREG/CR-4025, R&D Associates, August 1985 A-7 Severe Accident Risks: An Assessment for Five US Nuclear Power Plants, NUREG 1150, January 1991.
A-8 Technical Guidance for Siting Criteria Development, NUREG/CR-2239, Sandia National Laboratories, December 1982.
i S
i 53 Rev1
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2
25A5680 Table A 1 Radiological Consequences of ABWR Accident Sequences A
Whole Body Cumulative Exposure Probability Exposure,50 nele Risk Case (Event / year)*
(persorwem)
'(pernm/60 yr)
NCL 1.3E-07 9.60E3 0.075 1
2.1E-0G 1.38E4 0.017 2
7.8E-11 8.33E3 0.00004 3
0 3.71E5 0.000 4
0 2.06E5 0.000 5
7.5E-12 9.34E4 0.00004 6
3.1E-12 2.42E6 0.0004 7
3.9E-10 2.73E6 0.064 8
4.1 E-10 3.20E6 0.079 9
1.7E-10 3.31E6 0.034 Total:
0.269
- Sequences with probabilities of occurrence less than IE-9 per year are considered remote and speculative.
54 Rev1
c.
25A5680 Table A-2 Core Damage Frequency Contdbutors' Event Sequence Init.
Event 1A IBl IB2 IB3 ID II IIID IV Total Cont.
Scram 1 lE-08 4.3D10 9.5E-13 1.lE48 7.3 Turbine 6.8E-09 2.7L10 3.7&ll 7.lE49 4.5 Trip Isoladon 1.8508 7.lL10 1.lE-11 1.9E-08 11.9 LOOP 2 4.1509 1.5L11 4.2L13 4.lE-09 2.6 LOOP 8 2.4E-09 9.6612 1.4612 2.4LO9 1.5 LOOP 8+
5.8L10 1.lE-09 6.0Lil 1.7E-09 1.1 SHO2 6.6612 6.7E-08 6.7E-08 42.9 SBOB 2.6E 08 2.6L08 16.7 SBO8+
1.5E-08 8.9610 1.6E 08 10.3 IORV
- 1. l E-09 2.0E-10 9.5E-13 1.3E49 0.8 SB 2.5L10 2.5E-10 0.2 LOCA A'lWS 1.5E-10 1.5610 0.1 TOTAL 4.4608 2.6E-08 1.5E-08 8.9E-10 7.0E-08 1.lL10 2.5E-10 1.5E-10 1.57LO7 100 Offsite Release Group LCHP SBRC LCLP LilRC LBLC ATWS Total Case Case 1 3.4E-09 7.9L10 1.6E-08 5.lE-Il 2.0E 08 Case 2 7.8L11 7.8E-11 Case 3 1.3L12 1.3D12 Case 4 0
Case 5 6.3E-12 6.3L12 Case 6 1.2L10 1.2E-10 Case 7 1.1LIO 2.6L10 3.70L10 Case 8 2.1LIO 2.lE-10 Case 9 1.1E-12 1.5L10 1.5L10 NCL (N) 4.0E 08 1.5E-08 8.0E-08 2.0E-10 1.4E-07 Total 4.4E-08 1.6E48 9.6E 08 1.1 E-12 2.5L10 1.5E-10 1.57E 07 Contrib. %
28.1 10.3 61.4 0.122 0.2 0.1 100
- SAMDAs include both preventive and mitigative design alternatives 55 Rev1
_ c. -?. ;
?
25A5680 c
w.
Table A.S Modificadons Considered Modificadon Categosy -
i 1.
ACCIDENT MANAGEMENT
- a. Severe Accident EPGs/AMGs 2-
- b. Computer Aided Instrumentation 2
- c. Improved Maintenance Procedures / Manuals
.2
- d. Preventive Maintenance Features 4
- c. Improved Accident ManagementInstrumentation 4
- f. Remote Shutdown Station 1
- g. SecuritySystem I
- h. Simulator Training for Severe Accident 4
2.
REACTOR DECAY HEAT REMOVAL
- a. Passive High Pressure System 2
- b. Improved Depressurization 2
~!
- c. Suppression Pooljockey Pump 2
- d. Improved High Pressure Systems 1
i
- e. Additional Active High Pressure System 1
- f. Improved Low Pressure System (Firepump) 1
- g. Dedicated Suppression Pool Cooling 1
- h. Safety Related Condensate Storage Tank 2
i,16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> Station Blackout Injection 4
1
- j. Improved Recirculation Model 4
3.
CONTAINMENT CAPABILTIY
- a. LargerVolume Containment 2
- b. Increased Containment Pressure Capacity 2
- c. Improved Vacuum Breakers 2
- d. Increased Temperature Margin for Seals 1
- e. Improved Leak Detection 1
- f. Suppression Pool Scrubbing 1
- g. Improved Bottom Penetration Design 2
)
56 Rev1
p :.-..
25A5680 Table A.S Modificadons Considered Modificadon Categosy 1.
ACCIDENT MANAGEMENT
- a. Severe Accident EPGs/AMGs 2
- b. Computer Aided Instrumentation 2
- c. Improved Maintenance Procedures / Manuals 2
- d. Preventive Maintenance Features 4
- c. Improved Accident Management Instrumentation 4
- f. Remote Shutdown Station 1
- g. SecuritySystem I
h.' Simulator Training for Severe Accident 4
2.
REACTOR DECAY HEAT REMOVAL
- a. Passive High Pressure System 2
- b. Improved Depressurization 2
- c. Suppression PoolJockey Pump 2
- d. Improved High Pressure Systems I
- c. Additional Active High Pressure System 1
- f. Improved Low Pressure System (Firepump)
I
- g. Dedicated Suppression Pool Cooling I
- h. Safety Related Condensate Storage Tank 2
i.16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> Station Blackout Injection 4
- j. Improved Recirculation Model 4
3.
CONTAINMENT CAPABILITY
- a. LargerVolume Containment 2
- b. Increased Containment Pressure Capacity 2
- c. Improved Vacuum Breakers 2-
- d. Increased Temperature Margin for Seals 1
- c. Improved Leak Detection 1
- f. Suppression Pool Scrubbing 1
- g. Improved Bottom Penetration Design 2
56 Rev1
'J o n
,.e,
25A5680 i
~.:
4 Table A 3 (Condnued)
Modificadon Category i
4.
CONTAINMENT HEAT REMOVAL a.' Larger Volume Suppression Pool 2
- b. CUW Decay Heat Removal I
- c. High Flow Suppression Pool Cooling 1
- d. Passive Overpressure Relief I
i 5.
CONTAINMENT ATMOSPHERE MASS REMOVAL
- a. High Flow Unfiltered Vent 3
j
- b. High Flow Filte' red Vent 3
- c. Low Flow Vent (Filtered) 2 3
- d. Low Flow Vent (Unfiltered) 1 i
6.
COMBUSTIBLE GAS CONTROL
- a. Post Accident Inerting System 3
- b. Hydrogen Control byVenting-3
- c. Pre-inerting 1
- d. Ignition Systems 3
- c. Fire Suppression System Inerting 3
j i
7.
CONTAINMENT SPRAYSYSTEMS
- a. Drywell Head Flooding 2
- b. Containment Spray Augmentation 1
)
8.
PREVENTION CONCEI'rS
- a. Additional Service Water Pump 2
- b. Improved Operating Response 1
- c. Diverse Injection System 4
1
- d. Operating Experience Feedback 1
- e. Improved MSIV/SRV Design 1
9.
AC POWER SUPPLIES
- a. Steam Driven Turbine Generator 2
- b. Alternate Pump Power Source 2
- c. Deleted
- d. Additional Diesel Generator 1
57 Rev1
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L,*
25A5680 Table A-3 (Continued)
Modificadon Category 9.
(Continued)
- e. Increased Electrical Divisions 1
- f. Improved Uninterruptable Power Supplies 1
- g. AC Bus Cross-ties I
- h. Gas Turbine 1
- i. Dedicated RHR (bunkered) Power Supply 4
10.
DC POWER SUPPLIES
- a. Dedicated DC Power Supply 2
- b. Additional Batteries / Divisions 4
- c. Fuel Cells 4
- d. DC Cross-ties 1
- e. Extended Station Blackout Provisions 1
11.
ATWS CAPABILITY
- a. ATWS Sized Vent 2
- b. Improved ATWS Capability 1
12.
SEISMIC CAPABILFIY
- a. Increased Seismic Margins 1
- b. Integral Basemat 3
13.
SYSTEM SIMPLIFICATION
- a. Reactor Building Sprays 2
- b. System Simplification 1
- c. Reduction in Reactor Bldg Flooding 1
14.
CORE RETENTION DEVICES
- a. Flooded Rubble Bed 2
- b. Reactor Cavity Flooder 1
- c. Basaltic Cements 1
58 Rev1
=
,.yse j
l' 25A5680-Table A-4 Modifications Evaluated 1.
Accident Management la. Severe Accident EPGs/AMGs 1b. Computer Aided Instrumentation Ic. Improved Maintenance Procedures / Manuals 2.
Decay Heat Removal 2a. Passive High Pressure System -
2b. Improved Depressurization 2c. Suppression Pooljockey Pump 2d.. Safety Related Condensate Storage Tank 3.
Containment Capability Sa. Larger Volume Containment 3b. Increased Containment Pressure Capability i
Sc. Improved Vacuum Breakers 3d. Improved Bottom Head Penetration Design 4.
Containment Heat 4a. Larger Volume Suppression Pool I
Removal 5.
Containment Atmospbere 5a.
Low Flow Filtered Vent Gas Removal 7.
Containment Spray 7a. Drywell Head Flooding 8.
Prevention Concepts 8a. Additional Service Water Pump j
9.
AC Power Supplies 9a. Steam Driven Turbine Generator 9b. Alternate Pump Power Source
- 13. System Simplification 13a. Reactor Building Sprays
- 14. Core Retention Devices 14a. Flooded Rubble Bed 1
59 Rev1
,e w ss 25A5680 Table A-5 Summary of Benefits Averted Risk PotentialImprovement Person-rem la. Severe Accident EPGs/AMGs 1.5E-2 l b. Computer Aided Instrumentation 1.0E-2 I c.
Improved Maintenance Procedures / Manuals 1.6E-2 2a. Passive High Pressure System 6.9E-2 2b. Improved Depressurization 4.2E-2 2c. Suppression Pooljockey Pump 0.2E-2 2d. Safety Related Condensate Storage Tank 1.0E-2 Sa. Larger Volume Containment 15E-2 Sb. Increased Containment Pressure Capability 16E-2 Sc. Improved Vacuum Breakers 0.004E-2 3d. Improved Bottom Head Penetration Design 5.7E-2 4a. Larger Volume Suppression Pool 0.02E-2 5a. Low Flow Filtered Vent 1.4E-2 7a. Drywell Head Flooding 6.0E-2 i
8a. Additional Service Water Pump 1.6E-2 9a. Steam Driven Turbine Generator 5.2E-2 9b. Alternate Pump Power Source for high pressure systems 6.9E-2 10a. Dedicated DC Power Supply 6.9E-2 lla. ATWS Sized Vent 3.0E-2 1Sa. Reactor Building Sprays 1.7E-2 14a. Flooded Rubble Bed 0.1E-2 60 Revl
f t'
- n s~ v.
I' 25A5680 o
Taule A-6
&n===ry of Costs l
Esemated IWinimum Potentia! Improvement Cost -'
I a.
Severe Accident EPGs/AMGs
$ 600,000 lb. Computer Aided Instrumentation
$ 599,600 l
1c. Improved Maintenance Procedures / Manuals
$ 299,000.
2a. Passive High Pressure System
$ 1,744,000 2b. Improved Depressurization
$ 598,600
)
2c. Suppression PoolJockey Pump
$ 119,800 2d. Safety Related Condensate Storage Tank
$ 1,000,000 Sa. Larger Volume Containment
$ 8,000,000 Sb. Increased Containment Pressure Capability
$ 12,000,000 Sc. Improved Vacuum Breakers
$ 100,000 3d. Improved Bottom Head Penetration Design
$ 750,000 4a. Larger Volume Suppression Pool
$ 8,000,000 5a. Low Flow Filtered Vent
$ 3,000,000 l
7a. Drywell Head Flooding
$ 100,000 l
8a. Additional Service Water Pump
$ 5,999,000 9a. Steam Driven Turbine Generator
$ 5,994,300 9b. Alternate Pump Power Source
$ 1,194,000 10a. Dedicated DC Power Supply
$ 3,000,000 11a. ATWS Sized Vent
$ 300,000 13a. Reactor Building Sprays
$ 100,000 j
14a. Flooded Rubble Bed
$ 18,750,000 61 Revi l
n.
.n. - Lo 25A5680 l
Table A-7 Summary of Results j
Cost (K)/ Person-
]
Modification rem Averted j
7a.
Drywell Head Flooding
$1,667 13a. Reactor Building Sprays
$5,882 lla. ATWS Sized Vent
$10,000 3d.
Improved Bottom Penetration Design
$13,158 2b.
Improved Depressurization
$14,252 9b.
Alternate Pump Power Source f17,304 I c.
Improved Maintenance Procedures / Manuals
$18,688 2a.
Passive High Pressure System
$25,275 l a.
Severe Accident EPGs
$40,000 loa. De'dicated DC Power Supply
$43,478 3a.
Larger Volume Containment
$53,333 2c.
Suppression PoolJockey Pump
$59,990 l b.
Computer Aided Instrumentation
$59,960 Sb.
Increased Containment Pressure Capacity
$75,000 2d.
Safety Related Condensate Storage Tank
$100,000 9a.
Steam Driven Turbine Generator
$115,275 5a.
Low Flow Filtered Vent
$214,286 8a.
Additional Seivice Water Pump
$374,938 Sc.
Improved Vacuum Breakers
$2,500,000 14a. Flooded Rubble Bed
$18,750,000 4a.
Larger Volume Suppression Pool
$40,000,000 62 Revl FINAL
-