ML20081K911

Rev 0 to Technical Support Document for Abwr

ML20081K911
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Site:	05200001
Issue date:	11/30/1994
From:	GENERAL ELECTRIC CO.
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References
25A5680, 25A5680-R, 25A5680-R00, NUDOCS 9503290336
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1 TECHNICALSUPPORT DOCUMENT l l

FOR THE ABWR i

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1 General Electric Company SanJose, California November 1994 l

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p i TECHNICALSUPPORT DOCUMENT

. FOR THE ABWR TABLE OF CONTENTS ,

Section Iidt Sheet No.

EXECUTIVE

SUMMARY

4 l.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 I '/&fi,Su)N"('

POWER PLANTS 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 NORMAL OPERATIONS AND [gy y 5 v ;-"'k ' '-

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SEVERE ACCIDENTS IN PLANTS OF ABWR DESIGN 3.1 Radiological Risk from Normal Operadons of an ABWR Plant 11 3.2 Severe Accidents in Plants of ABWR Design .

12 3 3.3 Dominant Severe Accident Sequences for Plants of ABWR Design 14  :

3.4 Overall Conclusions from Chapter 19 of the ABWR SC.W. . 14 g'

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4.0 COST / BENEFIT EVALUATION OF SAMDAS FOR PLANTS OF ' y6!L/l , fi BWR DESIGN ' !'M 4.1 SAMDA Definition Applied to Plants of ABWR Design 15 4.2 Cost / Benefit Standard for Evaluation of ABWR SAMDAs 15 4.3 Candidate SAMDAs for the ABWR Design 15 4.4 Cost Estimates of Potential Modifications to the ABWR Design 16 4.5 Benefits of Potential Modifications to the ABWR Design 16 4.6 Cost / Benefit Comparison of SAMDAs 16 5.0

SUMMARY

AND CONCLUSIONS 17 l

6.0 REFERENCES

17 A'ITACHMENT A Evaluation of Potential Modifications to the ABWR Design 32 2 REV 0

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UST OFTABLES Inbla Ittis Sheet No.

1 Radiological Consequences of ABWR Accident Sequences 19 2 Severe Accident Mitigation Design Alternatives (SAMDAs)

Considered for the ABWR Design '20 3 SAMDAs Evaluated under NEPA for the ABWR 23 4 Cost Estimates of SAMDAs Evaluated for the ABWR under NEPA 26 5 Benefit Estimates of SAMDAs Evaluated for the ABWR under NEPA 28 6 Comparison of Estimated Costs and Benefits of SAMDAs ,

Evaluated for the ABWR under NEPA 30 A-1 Radiological Consequences of ABWR Accident Sequences 55 A-2 Core Damage Frequency Contributors 56 A-3 Modifications Considered 57 A-4 Modifications Evaluated 60 A-5 Summary of Benefits 61  !

A-6 Summary of Costs 62

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A-7 Summary of Results 63  ;

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EXECUTIVE

SUMMARY

The term " severe accident" refers to those events which are "beyond the substantial coverage of design basis events" and includes those for which there is substantial damage to the reactor coie 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).

For new reactor designs, such as the ABWR, the Nuclear Regulatory Commission (NRC), in satisfaction ofits severe accident safety requirements and guidance, 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 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 detennined that NEPA required consideration of certain design alternatives; namely, severe accident mitigation design alternatives (SAMDAs). See Limerick Ecoloey 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. Id. at 739.

Recently, the NRC Staff expanded the concept 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 policy.

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.

The 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; 4 REV 0

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(2) No cost-effective SAMDAs to the ABWR design have been identified to prevent or mitigate -

the consequences of a severe accident invohing 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 emironmental report, emironmental 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, ,

i (4) All reasonable features have been considered to reduce the radiological emironmental impacts from normal operations, including expected operational occurrences, and no further evaluation of such features or impacts shall be performed in any emironmental report, emironmental assessment, emironmental impact statement or other environmental analysis in connection with issuance of a license for a nuclear plant referencing a certified ABWR design.

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1.0 INTRODUCTION

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 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, 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 by limiting 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 t

.N safety requirements and policies. In particular, Chapter 19 includes: i

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(1) Identification of the dominant severe accident sequences and associated source terms for "i the ABWR design; .l (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 invohing substantial damage to the reactor core and to  ;

mitigate the consequences of such an accident should one occur," (Severe Accident Policy l 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 6 REV 0

~1 Consequently, the conclusions are drawn in Chapter 19 that further modifications to the ABh1 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 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 alternatives; namely, severe accident mitigation design alternatives (SAMDAs). Limerick Ecology 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 <ffective SAMDAs that should be added to these facilities ("NEPA/SAMDA 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.

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. I l

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 7 REV O

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f a  :: l Confidence Rule (10 CFR 851.23), 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 . l severe accident invohing substantial damage to the core and to mitigate the consequences of such an accident should one occur; (2) No cost-effective SAMDAs to the ABWR design have been identified to prevent or mitigate the consequences of a severe accident invohing 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 emironmental report, emironmental 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, (4) All reasonable features have been considered to reduce the radiological emironmental -

impacts from normal operations, including expected operational occurrences, and no further evaluation of such features or impacts shall be performed in any emironmental report, emironmental assessment, emironmental impact statement or other environmental i

analysis in connection with issuance of a license for a nuclear plant referencing a certified ABWR design.

The evaluation presented in this document is modeled after that found in the Limerick and i 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 f radiological risk from severe accidents is based on Chapter 19 of the ABWR SSAR. ppg e v 7A. ,

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• Description of Technical Support Document 1.3 .

Section 2.0 provides an oveniew of the radiological risks from normal operations and severe ,

accidents. Sections 3.0 through 5.0 provide the NEPA/SAMDA analysis. Section 3.0 discusses the methodological approach to the evaluation of SAMDAs 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.  ;

F 2.0 EVALUATIONS OF RADIOLOGICAL RISK FROM NUCLEAR POWER PLANTS 2.1 Evaluation of SAMDAs Under NEPA and Limerick Ecology Action Limerick Ecolocy 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 8 REV 0 l

remote and speculative. Second, the Commission may not make this finding and dispose of NEPA consideration oISAMDAs 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, ofwhich 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 evoludonary 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.

1 Chapter 19 of the ABWR SSAR demonstrates that these findings can be made for the ABWR t i

design. Given the nature and f'mdings of these severe accident and safety goal evaluations, GE believes that a sufficient basis exists for finding by rule that further consideration of severe accidents, including evaluation of SAMDAs pursuant to NEPA,is neither necessary nor reasonable.

2.2 Cost / Benefit Standard for NEPA Evaluation of SAMDAs The Limerick decision interpreted NEPA to require evaluation of SAMDAs for their risk reduction potential. In implementing the court's decision, the NRC consider:d the cost-effectiveness of each candidate SAMDA in mitigating the impact of a severe accident, using the

$1,000 per person-rem averted standard. This standard is a surrogate for all off-site 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 averted standard, which is what the Commission has historically used in distinguishing among and ranking design attematives, 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 10 CFR Part 50, Appendix I; the Safety Goal Policy Statement; the Severe Accident Policy Statement; and @50.34(f) requirements. In the context of l

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'I environmental analysis, it has been 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).

As indicated above, the Commission is preparing a Generic Environmental Impact Statement for 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). Additionally, Appendix I-j determinations are used to satisfy NEPA requirements with respect to radiological impacts from ,

normal operations.

.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.  ;

I 2.3 Socio-Economic Risks for Severe Accidents 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 SECY89-102, " Implementation of Safety Goal  ;

Policy." However, Environmental Impact Statements (EIS) for nuclear power plants provide separate, general discussions of the socio-economic risks from severe accidents. In keeping with i

this precedent, GE is providing a general discussion of socio-economic risks for the ABWR design, based in large measure on the discussion of such risks in NUREG-1437, " Generic Environmental Impact Statement for License Renewal of Nuclear Plants."

The term "socio economic risk from a severe accident" means the probability of a severe accident multiplied by the socio economic impacts of a severe accident. "Socio economic impacts," in turn, relate to off-site costs. The off-site costs considered in NUREG-1437 (see Vol. I, p. 5-90) are:

. Evacuation costs

. Value of crops or milk, contaminated and condemned l

. Costs of decontaminating property where practical

• Indirect costs due to the loss of the use of property or incomes derived therefrom (including '

interdiction to prevent human injury), and

• Impacts in wider regional markets and on sources of supply outside the contaminated are P, NUREG-1437 estimated the socio-economic risks from severe accidents. The estimates were 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. Higher 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 10-20% of off-site mitigation costs and were notincluded in the $6-58  :

billion dollar estimate. ,

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7 1Another 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 estimated to be small

. (10 acres / year at most). (See NUREG-1437, Vol. I, pp. 5-90 through 5-93.) .

i NUREG-1437 provides the bases for concluding that the socio<conomic risks from severe j accidents are predicted to be small and the residual impacts of severe accidents so minor that .

detailed consideration of mitigation alternatives is not warranted. See 56 Fed. Reg. 47,016, l 47,019,47,034 and 47,035 (September 17,1991).

The socio-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 NORMAL OPERATIONS AND SEVERE ACCIDENTS IN PIANTS OF ABWR DESIGN ,

3.1 Radiological Risk from Nonnal Operadons of an ABWR Plant

$50.34a and 650.36a of 10 CFR Part 50 require, in effect, that nuclear power reactors be designed [

and operated to keep levels of radioactive materials in gaseous and liquid effluents during normal  !

operations, including expected operational occurrences, "as low as reasonably achievable" (ALARA). Compliance with the guidelines in Appendix I to 10 CFR Part 50 is deemed a ,

conclusive showing of compliance with these ALARA requirements, j In addition to specif)ing numerical limits, Appendix I to 10 CFR Part 50 also requires an applicant to include in the radwaste system "all items of reasonably demonstrated technology that, when added to the system sequentially and in order of diminishing cost / benefit return, can, for a favorable cost / benefit ratio effect reductions in dose to the population reasonably expected  ;

to be within 50 miles of the reactor." The standard to be used in making this assessment is the cost / benefit ratio of $1,000 per person-rem averted.

The ABWR design complies with the guidance of Appendix I, as documented in Chapter 12 of the ABWR SSAR. Consequently, further consideration of alternatives to reduce the radiological i risks from normal operation of a plant of ABWR design is not warranted in order to satisfy NEPA.

Moreover, the radiologicalimpacts from normal operation of an ABWR are environmentally insignificant.

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c' Non radiological impacts from operation of an ABWR plant include those from the circulating system that removes heat from the reactor (e.g., cooling towers, cooling lakes, etc.), intake systems for the water in the circulating systems, discharge systems for the water in the circulating system, biocide treatment in circulating water to prevent fouling by organisms, chemical waste treatment and disposal, sanitay waste treatment system, and electrical transmission facilities.

Each of these systems is part of that portion of the ABWR design that is not being certified because it is site-specific. It may be appropriate to consider design alternatives for non-radiological systems under NEPA. However, the choice of alternative will not have an effect on the portion of the ABWR design that is being certified. Consideration of alternative designs to systems affecting non-radiological impacts must be done on a site-specific basis.

3.2 Severe Accidents in Plants of ABWR Design Chapter 19 of the ABWR SSAR, " Response to Severe Accident Policy Statement," establishes that /f the Commission's severe accident safety requirements have been met for the ABWR design, ;i 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.

Specifically, the following topics were addressed in Chapter 19 of the ABWR SSAR:

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(1) Consideration of the contributions ofinternal events (Section 19.3), Shutdown events (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 reduction features which were included in the ABWR design to achieve accident prevention and mitigation (addressed in Subsection 19.7.3(2));

'M T Consideration of additional modifications, ev, aluated in accordance with $50.34(0(1), is _..

addressed in Attachment A.l Attachmi:nt Alon~cl6 des ~tliaithe severe aTcident reqmrements of 10, CFR Part 52 (@523f(a)(1)(ii), (iv) & (v)) and the Severe Accident Policy Statement have been. ,

met. It also p_rovides a summary of the bases for these conclusions.% articular, Attachment A

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presents a summag of the bMes for coliEluding that the requireinents of $50.34(0 (referenced in 552.47(a)(1)(ii)) have been met, including $50.34(0 (1)(i), which requires " perform [ance of] a plant / site-specific [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 notimpact excessively on the plant." Attachment A also 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.

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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:

• Prevention of core damage

• Prevention of early containment failure for dominant accident sequences

• Evaluation of the effects of hydrogen generation

• 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 p! ants of ABWR design based on the ABWR !i SSAR Chapter 19 evaluations are as follows:

(1) Core Damage Frecuency. The ABWR core damage frequencywas determined to be 1.6E-7 per reactor year in Subsection 19.6.2. The goal was 1E-6 per reactor year.

(2) Conditional Containment Failure Probability. 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) Individual Risk (Prompt Fatality Risk). 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 goal is 3.9E-7 per individual per year (or 0.04 per 100,000 people per year).

(4) Societal Risk (Latent Fatality 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 significantlyless 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 1.7E-6 per individual per year (or 0.17 deaths per 100,000 people per year).

(5) Probability of12rge Off-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 , 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.

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-l l 3.3 ~ 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. l 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 VermontYankee Nuclear Power Corooration, (Vermont Yankee Nuclear Power Station), CLI-9007,32 NRC 129,132 (1990).

GE believes that the severe accident analysis in Chapter 19 of the ABWR SSAR provides a i sufficient basis for the Commission to find that ABWR sequences that are not dominant can be deemed remote and speculative. - .

t 3.4 Overall Conclusions from Chapter 19 of the ABWR SSAR l'

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-effective modifications to the ABWR design have been identified to reduce the risk from a severe accident invohing sul ,cmtial 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.

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• COST / BENEFIT EVALUATION OF SAMDAS FOR PIANTS OF ABWR DESIGN:

(4.1 SAMDA Dennidon Applied to Plants of ABWR Design

. Attachment A' considers whether the ABWR design should be modified in order to prevent orL 1 mitigate the consequences of a severe accidentin 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 1457: design alterna:ives that could prevent and/or mitigate the consequences of a severe accident. ,

4.2 Cost / Benefit Standard for Evaluadon 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 l evaluating SAMDAs under NEPA. This standard was used as a surrogate for all off-site cosu 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 nature'. 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 imnlementation MINUS averted on-site costs

• Reduction in residua'l risk (person-rem /plantlife) j A plant lifetime of 60 years was assumed to maximize the reduction in residual risk. .

4.3 Candidate SAMDAs for the ABWR Design The complete list of SAMDAs considered for plants.of ABWR design is contained in Table 2. The SAMDAs are classified according to the following categories:

(1) Modification is applicable to the ABWR and alreadyincorporated 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.

' 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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(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 Attachment A.

4.4 Cost Estimates of Potential Modificadons 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 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 est: mated costs of design alternatives that are, at least partially, preventive in nature were adjusted thr 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. 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 modificadan. Actualimplementation costs are expected to be significantly higher than those used in this evaluation.

4.5 Benefits of Potential Modifications to the ABWR Design Table 5 summarizes the basis for assigning a benefit to each SAMDA. In general, benefits were l estimated from the PRA results of Chapter 19 of the ABWRSSAR by considering which sequences '

are affected by each modification. Detailed discussion of the method for estimating benefit is provided in 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 i evaluation since the cost / benefit standard was not met. The closest design alternative exceeds ,

l the criteria by more than a factor of 1000.

l 1

l 16 REV O

\

, i On th'e 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. ~

'l

1 5.0

SUMMARY

_ AND CONCLUSIONS )

J 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 a criterion of $1,000 per person-rem averted was used to determine which alternatives, if any, w re  :

< cost-effective. 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 ofi j the ABWR design,' SAMDAs cannot be reasonably incorporated in a cost-effective manner.  !

J 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. -l 1

6.0- REFERENCES  !

va

1. 'ABWR Standard Safety Analysis Report,23A6100, Docket No.52-001, GE Nuclear Energy. ) [D  ;

2. Assessment of Severe' Accident Prevention and Mitigation Features, NUREG/CR-4920, Brookhaven National Laboratory, July 1988.  ;

3. Design and Feasibility of Accident Mitigation Systems for Light Water Reactors, ,

NUREG/CR-4025, R&D Associates, August 1985.

i 4.. Evaluation of Proposed Modifications to the GESSAR II Design, NEDE 30640 (Proprietary),  ;

June 1984.

5. Generic Emironmental Impact Statement for License Renewal of Nuclear Plants, NUREG- l 1437, August 1991. ,

6. " Issuance of Supplement. to the Final Emironmental Statement-Comanche Peak Steam  ;

Electric Station, Units i and 2", NUREG 0775 Supplement, December 15,1989.  !

7. Severe Accident Risks: An Assessment for Five US Nuclear Power Plants, NUREG 1150,  ;

Januaq1991. l

8. " Supplement to the Final Emironmental Statement-Limerick Generating Station, Units 1 and 2", NUREG 0974 Supplement, August 16,1989. j W

17 REV 0 .

e e g es m - e --.

l

- 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.

t i

j i

a 18 REV 0 I l

1

1 l l i

Table 1 Radiological Consequences of ABWR Accident Sequences Whole Body Cumuladve Exposure Probability Exposure,50 mile Risk Case (Event / Year)* (Person-rem) (Per-rem /60 Yr)

NCL 1.3E47 9.60E3 0.075 1 2.1E48 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.

(cwsr.v.OD;- N (fgout v- 7) f3. 2 %/0 P. Adh >( W. ! y/[ fyg) X H '/M 5 GVGHT 7,9 X /0 P- RCn1 Geo W S 19 REV 0

o j t

Table 2 Severe Accident Midgation Design Alternatives (SAMDAs)*

Considered for the ABWR Design Modification Category

1. ACCIDENT MANAGEMENT

a. Severe AccidentEPGs/AMGs 2

b. Computer AidedInstrumentation 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 DECAYHEAT REMOVAL

a. Passive High Pressure System 2 ,

b: Improved Depressurization 2

c. Suppression PoolJockey Pump 2

d. Improved High Pressure Systems 1

e. Additional Active High Pressure System 1

f. Improved Low Pressure System (Firepump) I

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

j. Improved Recirculation Model 4

3. CONTAINMENT CAPABILITY a.12rger 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 20 REV 0

I

l l I

. ' ~ .

1 Table 2 (Continued) i Modification Category

4. CONTAINMENT HEAT REMOVAL l

a. LargerVolume Suppression Pool 2

b. CUW Decay Heat Removal 1.

c. High Flow Suppression Pool Cooling 1-

d. Passive Overpressure Relief 1 i

5. CONTAINMENT ATMOSPHERE MASS REMOVAL

a. High Flow Unfiltered Vent 3 ,

b. High Flow Filtered Vent 3

c. Low Flow Vent (Filtered) 2

d. Low FlowVent (Unfiltered) .I

6. COMBUSTIBLE GAS CONTROL

a. Post AccidentInerting System 3

b. Hydrogen Control byVenting 3

c. Pre-inerting I

d. Ignition Systems 3

c. Fire Suppression System Inerting 3

7. CONTAINMENT SPRAYSYSTEMS

a. Drywell Head Flooding 2 -i

b. Containment Spray Augmentation 1

8. PREVENTION CONCEPTS

a. Additional Service Water Pump 2

b. Improved Operating Response 1

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. Addidonal Diesel Generator 1 21 REV O

l =

Table 2 (Continued)

Modification Category

9. (Continued)

c. bcreased Electrical Divisions 1

f. Impoved Uninterruptable Power Supplies 1

g. AC Bus Cross-ties I

h. Gas Turbine 1

i. Dedicated RHR (bunkered) Power Supply 4

10. DC POWER SUPPI.lES

a. Dedicated DC Power Supply 2

b. AdditionalBatteries/ 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 EISMIC CAPABILITY

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 I

I 22 REV 0

- = . -

Table 3 SAMDAs Evaluated Under NEPA for the ABWR PotentialImprovement Advantages Disadvantages la. Severe Accident Improved arrest of core melt None EPGs/AMGs progress and prevention of containment failure.

Ib. Computer Aided Improved prevemion ofcore Additional training Instrumentation melt sequences Ic. Improved Maintenance Improved prevention ofcore Increased documentation cost Procedures / Manuals melt sequences 2a. Passive High Pressure Improved prevention ofcore High cost of additional system System melt sequences 2b. Improved Depressurization Improved utilization ofLow Cost of additional equipment Pressure sy~ ems for prevention of core melt sequences 2c. Suppression Pool Jockey Improved prevention of core Cost of additional equipment Pump melt sequences 2d Safety Related Condensate Availability following Seismic Design and structural costs Storage Tank events 3a. Larger Volume Containment a. Increases time before a. High cost (Double Free Volume) containment failure b. Containment failure not

b. Increases time for recovery prevented

c. Minor radiological benefit since risks dominated by long lived isotopes 3b. Increased Containment a. Eh. inates large releases a. Extreme costs  :

Pressure Capability b. High temperature failures (Sufficient pressure to not prevented  ;

withstand severe accidents) 3c. Improved Vacuum Breakers a. Reduces probability of a. Increased maintenance and ,

(Redundant valves in each suppression pool bypass equipment costs t line) 3d. Improved Bottom Head a. Increased time for in-vessel a. Cost for equipment and  !

Penetration Design arrest analysis 4a. Larger Volume Suppression a. Increases heat absorption a. High cost Pool (Double effective liquid capability within volume) containment 23 REV 0 w

, Table 3 (Continued)

PotentialImprovement Advantages Disadvantages 4a. (Continued) b. Increases time for recovery b. Minor radiological benefit ofsystems since risks dominated by

c. Increases time before long lived isotopes containment failure Sa. Low Flow Filtered Vent a. Provides some scrubbing of a. Probability of drywell head fission products ifhead fails failure is low relative to the

b. Reduces containment other containment failure leakage ifmovable modes penetrations are 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 lla. ATWS Sized Vent a. Provides scrubbing of a. Uncertain location 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 Sssion Uncertain location and (Firewater crosstie for products from Reactor unknown potential reactor building sprays) Building consequences from inadvertent actuation 24 REV 0

Table 3 (Contioned) i PotentialImprovement Advantages Disadvantages i < ,1 14a. Flooded Rubble Bed Prevention ofcore-concrete Small benefit over passive k . ,.h interaction afects . flooding system. k'

,j'-

T l

1 t

25 REV 0

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 5 600,000 EPGs/AMGs beyond generic work by Owners' Group.

I b. Computer Aided Software modifications and interface $ 599,600 l Instrumentation hardware. Credit for averted onsite cost included.

Ic. Improved Maintenance Procedur: 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 Logic, pneumatic supplies, piping and 5 598,600 Depressurization 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 Structural analysis and material. Credit for 5 1,000,000 Condensate Storage Tank averted onsite cost included.

Sa. Larger Volume Double currentvolume at $1200/ft'. $ 8,000,000 Containment (Double Free Analysis not included.

Volume)

Sb. Increased Containment Similar to Larger Volume Containment, $ 12,000,000 Pressure Capibility but denser rebar andlabor required.

(Suflicient pressure to Assumed 50% higher cost.

withstand severe accidents)

Sc. Improved Vacuum Eight lines at $10,000 per line. $ 100,000 Breakers (Redundant valves in each line) 26 REV 0

Table 4 (Continued)

Potential Estimated Improvement Cost Basis M'mimum Cost 3d. Improved Bottom Head 205 drives at $1,000/ drive and $500,000 of $ 750,000 Penetration Design analysis 4a. Larger Volume Assitmed to be the same as Larger Volume $ 8,000,000 Suppression Pool (Double Containment effective liquid volume)

Sa. Low Flow Filtered Vent Hardware and Testing program $ 3,000,000 7a. Drywell Head Flooding Minor valve and piping modification with $ 100,000 (Firewater crosstic to instrumentation drywell head area) 8a. Additional Senice Water System hardware, power suplies and S 5,999,000 Pump support systems. Credit for averted onsite cost included.

9a. Steam Driven Turbine System hardware, cabling and structural $ 5,994,300 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.

10a. Dedicated DC Power 5000 ft' building structure addition at $ 3,000,000 Supply $500/ft' and csbling 1la. ATWS Sized Vent Instrumentation and cabling in addition to $ 300,000  !

training  !

13a. Reactor Building Sprays Minor valve and piping modification with $ 100,000 I (Firewater crosstic for instrumentation.

reactor building sprays) 14a Flooded Rubble Bed 1250 ft' of material at $1000/lb $ 18,750,000 I

4 27 REV O

i .'

l l

~

Table 5 j Benefit Estimates of SAMDAs* i Evaluated for the ABWR Under NEPA i

Potential Averted Risk Improvement Benefit Basis Person-rem l a. Severe Accident 10% improvement in mitigation actions 0.015 EPGs/AhfGs Ib. Computer Aided 10% improvement in preventative actions 0.01 Instrumentation Ic. Improved hiaintenance 10% improvement in reliability of RCIC, 0.016 Procedures /hfanuals HPCF, RHR and LPFL 2a. Passive High Pressure 90% reliable diverse additional high 0.069 System pressure system 2b. Improved 50% reduction in manual depressurization 0.042 Depressurization reliability  :

2c. Suppression PoolJockey 10% improvement in low pressure makeup 0.002 Pump reliability.

2d. Safety Related Arbitrary selection due to high suppression 0.01 Condensate Storage Tank pool availability.

Sa. Larger Volume Elimination of drywell head failure 0.15 Containment (Double Free sequences Volume) 3b. Increased Containment Elimination of all cases except normal 0.16 Pressure Capability containmentleakage (NCL)

(Sufficient pressure to withstand severe accidents)

Sc. Improved Vacuum Elimination of Case 2 sequences 0.00004 Breakers (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 Elimination of Case 9 sequences involving 0.0002 Suppression Pool (Double loss of suppression pool cooling systems effective liquid volume)

• SAhfDAs include both preventive and mitigative design alternatives 28 REV 0

l Table 5 (Continued)

Potential Averted Risk Improvement Benefit Basis Person rem 5a. Low Flow Filtered Vent Elimination of sequences involving 0.014 initiation of containment rupture disc 7a. Drywell Head Flooding Reduction in high temperature 0.06 (Firewater crosstic to containment failure sequences and drywell drywell head area) head failure sequences Ba. 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 system. 0.069 Source for high pressure See 2a.

systems 10a. Dedicated DC Power Similar to additional high pressure system. 0.069 Supply See 2a.

lla. ATWS Sized Vent Reduction in Case 9 sequences 0.03 13a Reactor Building Sprays 10% reduction in consequence of 0.017 (Firewater crosstic for sequences invohing containment leakage reactor building sprays) 14a. Flooded Rubble Bed Elimination of sequences involving core- 0.001 Concrete interaction 29 REV 0

i i

~

Table 6 Comparison of Estimated Costs and Benefits on SAMDAs*

Evaluated for the ABWR Under NEPA Cost. Benefit l Estimated Ratio I Minimum Cost Averted Risk ($K per Person.

PotentialImprovement ($) Person-rem rem) la. Severe AccidentEPGs/AMGs $ 600,000 0.015 S 40,000 lb. Computer Aided Instrumentation 5 599,600 0.01 S 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 Poo'. Jockey Pump $ 119,800 0.002 5 59,900 2d. Safety Related Condensate Storage $ 1,000,000 0.01 S 100,000 Tank Sa. Larger Volume Containment S 8,000,000 0.15 5 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 S 2,500,000 (Redundant valves in each line) 3d. Improved Bottom Head Penetration S 750,000 0.057 5 13,160 Design 4a. Larger Volume Suppression Pool $ 8,000,000 0.0002 3 40,000,000 (Double efTective liquid volume) 5a. Low Flow Filtered Vent $ 3,000,000 0.014 $ 214,300

~

7a. Drywell Head Flooding (Firewater 5 100,000 0.06 5 1,700 crosstic to drywell head area)

• SAMDAs include both preventive and mitigative design alternatives 30 REV 0

~

Table 6 (Continued)

Cost-Benefit Estimated Ratio Minimum Cost Averted RlA ($Kper Person-Potential Improvement ($) 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 SizedVent $ 300,000 0.03 $ 10,000 13a. Reactor Building Sprays (Firewater $ 100,000 0.017 $ 5,900 crosstic for reactor building sprays) 14a. Flooded Rubble Bed l $ 18,750,000 0.001 $ 18,750,000 4

a 31 REV 0

XITACHMENT 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 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 of 10CFR50.34(f). Item (f)(1)(i) requires performance of a plant site-specific [PRA] the aim ofwhich 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 prosides the base PRA of the ABWR plant.

To address this requirement, a review of potendal modifications to the ABWR design, beyond ,

those included in the Probabilistic Risk Assessment (PRA), was conducted to evaluate whether 7' potential severe accident design features could bejustified on thgasis of cost per person-Sievert 0 averted. ~-- ~

w ,

This attachment summarizes the results of GE's review and evaluation of the AB%R 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 attraClive.

A.I.2 Evaluation Cdteria 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 (n.persongievert) 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. Consistentwith 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.

• Attachment A is updated version of ABWR SSAR Appendix 19P of the same title.

32 REV O

decontamination of contaminated land w .

n s, elimination ofland use and considered in this evaluation ascation. rectcredits against in accident costs are events which have high consequence. )The maximu sults from verylow probability determined to be $269. Therefore, based on this methodology cost of However, potential a variety of modifications were reviewed to establish th, no modi changes.

e relative attractiveness of A.I.3 Methodology The overall approach was to estimate the benefit ofmodificati tb favor nib ~difications. Because of the uncertaintiperson-Sie s were assessed in order this study. severe accidents with sensible modifications, this basis isjudgedes to be acceptable for purposes of t

A.I.3.1 Selection of Modifications

{

{ Potential modifications were identified from a variety of previo

{ studies of preventative and mitigative features which address severus industr 1

selected for further review based on ,

beingcomposite list ofmo ential modifications were

( (1) i applicable to the ABWR design, and (2) not included in the reference PRA.

Additional detail on the selection of modifications is provided i S n ection A.S.

A.1.3.2 Costs Basis Rough order of magnitude costs were assigned for each modifi incremental costs that would be incurresptems and system improveme represent the estimated backfit basis. Section A.5 defines the cost estimates for each of thn a new pla e modifications.

Even for a new plant such as the ABWR, relatively large costs (se arrangement. This is because the e u cost oflabor ing structures or and mater area required. For other modifications which involve minor hardwarn a function of th e addition, the cost is often 33 RE

~

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 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 include replacement power 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 presentworth of the time dependent economic risks.

(2) Replacement power was based on a rate of $.013/kW-h differential as bar cost. The differential rate was assumed to be constant over the remaining life of the plant.

(3) The economic ulue 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 facilitywere assumed to be $2 Billion.

(5) The economic evaluations were based on a discount rate of 8% and escalation factor of 3%.

A.I.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 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.

34 REV C j

' Sectio'n A.4 summarizes each concept and estimated benefit for each individual potential modification. For each modification the cost per person 4igvert averted was evaluated to obtain

• the results of the individual evaluations. These conds are provided in Section A.7.

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.

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 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.

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 i the current ABWR design-specifically to address severe accidents. These features and the  !

reference description include, f

1 l

Design Feature SSAR References l l

(1) Firewater pump crosstie 5.4.7.1.1.10 (2) Passive containment flooder 9.5.12 (3) Gas turbine generator 9.5.11 l (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 oiTsite 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:

i l

l l

1 l

35 REV 0 j l

,- 'q .'.. ,

i Case '1 . Core Melt arrested in vessel or in Containment with actuation of containment rupture disk.  ;

a.

Case 2 ~ Low Pressure Core Melt with suppression pool bypass and actuation of containment nipture disk.  !

-Case 3 - High Pressure Core Melt with drywell Head failure and fire water spray initiation.

' Case 4 _ Supprenion 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 initiation.

Case 7 . Low Pressure Core Melt with Drywell Head failure and no mitigation Case 8 High Pressure Core Melt with Early Containment failure.

Case 9 A'IWS event with Drywell Head failure.

NCL Normal Containment Leakage to Reactor Building.- i The offsite exposures for each case shown in Table A-1 were calculated by the CRAC2 code for five representative US regions for t!.e 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 by low pressure transient events (LCLP) (61.4%), followed by high pressure transient events (LCHP) (28.1%) 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.- i 1

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 19.7-36 REV 0

-~

n

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:

(1) Modification is applicable to ABWR and already incorporated 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 benefits of potential modifications to the ABWR design identified in Table A-4. For each modification the basis for the evaluation and the conceptis described. Table A-5 summarizes the benefit in terms of person-Sievert 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.4.1.1 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 (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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l 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 care melt arrest 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 already part of the APWR control room design. Additional artificialintelligence 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 incremental benefit over severe accident EPGs (Subsection A.4.1.1) is about 3% in core damage frequency (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 additionalinformation 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 different degrees.

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 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 38 REV C

. ll ; l 4

~

(NUREG 0737, II.K.3.15). Additionally, the redundant HPCF is an improvement over early - '

productlines which used the single HPCF system.

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 tystem to that under consideration for the  :

' Simplified BWR (SBWR) is employed. j 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.-

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

. factor of 2. Based on this estimate of benefit offsite person-Sievert 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 i

return path to the suppression pool would be through existing piping such as shutdown cooling

- return lines.

6 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 j 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 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 itwould be 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 19I 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 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.

A.4.3.1 Larger Volume 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 containment failure.

This item would mitigate the consequence of an accident by delaying the time before the severe 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 fail j due to the long term pressurization caused by core concrete interaction and steam generation. ,

1 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 The design pressure of the ABWR containmenti' 45 psig, The containment rupture disk pressure and ultimate capability are significantly L;her. by increasing the ultimate pressure 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.

This modification would mitigate the event, not change the core damage frequency and the increased pressure capability may not be sufficient to contain the long term pressurization caused by core concrete interaction and steam generation. However, ifit were able to prevent all severe source term release except for normal containment leakage, the person-Sievert risk would be about 0.02 person-rem /60 years. Therefore, the benefit would be about 0.16 person-rem.

A.4.3.3 Improved Vacuum Breakers 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 eliminate the need for operator action.

If Case 2 sequences were climinated, the benefit of this modification would be about 0.00004 person-rem averted.

L ,

t A.4.3.4 Improved Bottom Head Penetration Design ,' .-

p 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 bottom head by the CUW system. A carbon steel transition piece connects the drain line to the RPV. During a severe accident this transition piece may be susceptible to melting and may 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 internal 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 containment.

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 1

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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 welds. The concept would be to make such external welds and supports small enough that the 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 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 The ABWR design contains 3 divisions of suppression pool cooling and provisions for a 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 offsite exposures. Additional modifications are notlikely to show substantial safety benefits.

A.4.4.1 Iarger Volume Suppression Pool This item would increase the size of the supptession 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 benefit would be elimination 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. j The assessed maximum benefit is therefore about 0.0002 person-rem.

A.4.5 Containment Atmosphere Mass Removal l

l The ABWR design contains a containment rupture disk which provides contamment overpressure l 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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A.4.5.1 Low Flow 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 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 DnwellHead Flooding This concept would provide intentional flooding of the upper drywell head such that if high dqwell temperatures occurred, the drywell head seal would not fail. Additionally,if the seal were to fail due to overpressurization of the dgwell, some scrubbing of the released fission products would occur. This system would be designed to operate passively or use an AC-independent water source.

If an extension of the fire pump to dgwell spray crosstie were considered for manualinitiation 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 person-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 43 REV 0

.- 1 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

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 water systems, could conceivably reduce the plant risk through an improvement in system reliability. The 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 somewhatless 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.

44 REV 0

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 PRA indicates 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 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 injection 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.

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 45 REV 0

~  !

contiibute about 0.1% of the core damage frequency (Table A-2) and about 17% of the offsite i risk (Case 9).  ;

A.4.11.1 ATWS Sized Vent j 1

This modification would be available to remove reactor heat from ATWS events in addition to severe accidents and Class II 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 19I of the ABWR SSAR) addresses the margins associated with the -

seismic design and concludes that there is a 95% confidence that existing 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 addidonal changes are considered.

i 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 seismic and pipe whip restraints is included in the concept.

While there are several examples of redundant sptems, 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.

46 REV 0

,~ [ l l

The benefit of this modification could be to reduce the impact of events which do not involve the I operation of the containment rupture disk. Such events relene fission products from the l containment into the reactor building. Releases from normal containment leakage and cases 3, 6,7,8 and case 9 sequences could potentially be reduced. If10% of these releases from these cases were arbitrarily mitigated by this method, the benefit would be about 1.7E-04 person-S: evert.

i I

i

( A.4.14 Core Retention Devices l

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 overflowinto the lower drywell after a few hours and quench the l debris bed. After the molten core has been quenched, no further ablation of concrete is j expected and the decay heat can be removed by normal containment 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 additional concept from Reference A-4 is included.

A.4.14.1 Flooded 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 (ThO2) pellets in a multiple layer geometry has been shown to stop melt penetration; thus, preventing core-concrete 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 than the corium. A major drawback common to all flooded rubble bed core retention systems is the need for further experimental testing in order to 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 of the ABWR SSAR indicates a 90% certainty that debris on a concrete floor covered with water will be coolable in the current ABWR design.

Only sequences in which no liquid injection to the drywell occurs will result in core-concr ete interaction. A conservative estimate of the benefit of this concept over the existing design would 47 REV 0

be e'limination of sequences with core-concrete interaction except those with containment

. cooling failure. A review of Subsection 19E.2 of the ABWR SSAR indicates that this would effect about 1% of Cases 1,6 and 7. This corresponds to about 0.001 person-rem averted.

A.5 COST IMPACTS OF POTENTIAL MODIFICATIONS 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 modificadon are discussed in Subsection A.1.3.2. 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 at least $600,000 for plant specific modifications to EOPs.

A.5.1.2 Computer Alded Instmmentation 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 modification reduces the frequency of core damage events, a present worth of $400 onsite costs are averted and the cost basis is $599,600.

i 48 REV O

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- A.5.1.3 Improved Maintenance Procedures / Manuals The cost of at least $300,000 would be required to identify components which should receive - l enhanced maintenance attention and to prepare the additional detailed procedures or, recommended information beyond that currently planned. Credit for reduction in onsite costs reduces the cost basis to $299,000.

l 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 it would 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 ABWR design. The cost 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 a building space are not required. Controls and associated wiring could cost an additional $60,000 for a total cost of atleast $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 Imger 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 not include 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 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 atleast $12,000,000 could be required. Since this ,

modification is mitigative, no credit for onsite averted costs was assumed.

l A.5.3.3 Improved Vacuum Breakers The cost of redundant vacuum breakers including installation and hardware is estimated at more than $10,000 per line. Instrumentation associated with this modification is notincluded. 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 l

The cost increase of using a stainless or inconel transition piece as opposed to carbon steel would be expected to be smallin comparison to the engmeering and documentation change costs associated with the change. Costs, associated with external welds and support for the CRDs is judged to be at least $1000 per drive. In addition, about $500,000 of analysis would be required 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 mitiga:ive, 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 contr.inment. An estimate of $8,000,000 is assigned to this item. I 50 REV O l

. j. l m i A.5.51 Containment Atmosphere Mass Removal  !

i i

- A.5.5.1 Low Flow Mkered Vent

~ 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. <

t Since this modification is mitigative, no credit for averted onsite costs applies.

1 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 Drywell Head Flooding c

An additionalline to flood the drywell head using existing Jrewater piping would be a relatively '

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. {

t A.5.8 Prevention Concepts ,

A.5.8.1 Additional Service Water Pump l The use of diverse instrumentat% v ould not presumably have a significant equipment cost, but there would be an' increased cost d inaintenance and spare parts due to less interchangeability and less standardization of procedures. ,

1 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 atleast $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.4.13 and A.5.1.3.  !

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A.5.9 AC Power Supplies A.5.9.1 Steam-Driven Turbine Generator 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 t

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 i 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 l 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 l would be required for DC pumps, cabling and instrumentation and controllers. A total cost l would be at least $3,000,000. j I

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A.5.I1 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 e 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.

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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 j 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.

Based on this low level no modifications arejustified for the ABWR. Based on the PRA results, none of the modifications provided a substantialimprovement in plant safety.

A.8 REFERENCES A-1 Evaluation of Proposed Modifications to the GESSAR II Design, NEDE 30640 (Proprietary), June 1984.

A2 Supplement to the Final Environmental Statement - Limerick Generating Station, Units 1 and 2. NUREG4974 Supplement, August 16,1989 A3 Issuance of Supplement to the Final Emironmental Statement- Comanche Peak Steam Electric Station, Units 1 and 2, NUREG 0775 Supplement, December 15,1989 A-4 Survey 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, NUREG/CR-4025, R&D Associates, August 1985 A7 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.

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Table A-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 0.0004 7 3.9E-10 2.73E6 0.064 8 4.1E-10 3.20E6 '.079 0

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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Table A-2 Core Damage Frequency Contributors

• l Event Sequence  ;

Init. I  % I Event 1A IBl IB2 IB3 ID II IIID IV Total Cont.

Scram 1.l E-08 4.3E-10 9.5E-13 1.lE-08 7.3 Turbine 6.8E-09 2.7E-10 S.7E-Il 7.lE-09 4.5 Taip Isolation 1.8E-08 7.lE-10 1.lE-Il 1.9E-08 11.9 LOOP 2 4.lE-09 1.5E-11 4.2E-13 4.lE-09 2.6 I LOOPS 2.4E 09 9.6E-12 1.4E-12 2.4E49 1.5 LOOP 8+ 5.8E-10 1.lE 09 6.0E-Il 1.7E49 1.1 SB02 6.6E 12 6.7E-08 6.7E-08 42.9 SBO8 2.6E48 2.6E-08 16.7 SBO8+ 1.5E 08 8.9E-10 1.6E-08 10.3 IORV 1.l E-09 2.0E-10 9.5E-13 1.3E-09 0.8 ,

SB 2.5E-10 2.5E-10 0.2 LOCA A7WS 1.5E-10 1.5E 10 0.1 TOTAL 4.4E-08 2.6E-08 1.5E 08 8.9E-10 7.0E-08 1.lE-10 2.5E-10 1.5E 10 1.57E-07 100, Offsite Release Group LCHP SBRC LCLP LHRC LBLC ATWS Total Case Case 1 3.4E-09 7.9E-10 1.6E-08 5.lE Il 2.0E-08 Case 2 7.8E-11 7.8E 11 Case 3 1.3E 12 1.3E 12 Case 4 b Case 5 6.3E-12 13E12 Case 6 1.2E-10 1.2E-10 Case 7 1.1E 10 2.6E 10 3.70E 10 Case 8 2.lE 10 2.1E 10 Case 9 1.lE-12 1.5E 10 1.5E-10 NCL (N) 4.0E 08 1.5E-08 8.0E-0F 2.0E-10 1.4 E-07 Total 4.4E 08 1.6E-08 9.6E-08 1.lE-12 2.5E-10 1.5E-10 1.57E-07 Contnb. % 28.1 10.3 61.4 0.122 0.2 0.1 100 i

• SAMDAs include both preventive and mitigative design alternatives 56 REV 0

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Table A-3 Modifications Considered Modification 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

e. Improved Accident Management Instrumentation 4

f. Remote Shutdown Station I

g. Security System 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 Pool Jockey Pump 2

d. Improved High Pressure Systems 1

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 lajection 4

j. Improved Recirculation Model 4

3. CONTAINMENT CAPABILITY

a. Larger Volume Containment 2

b. Increased Containment Pressure Capacity 2

, c. Improved Vacuum Breakers 2

d. Increased Temperature Margin for Seals I

e. Improved Leak Detection 1

f. Suppression Pool Scrubbing 1

g. Improved Bottom Penetration Design 2 i

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_ ,_ _ __ . . _ __ . __ -. .-_,m 7

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.l ; . ': .f ,i I- Table A-3'(Continued)  !

Modification Category

{

4. - CONTAINMENT HEAT REMOVAL  !

' a. Larger Volume Suppression Poct 2

b. CUW Decay Heat Removal 1

c. High Flow Suppression Pool Cooling- 1 l i

d. Passiva Overpressure Relief 1

)

5. CONTAINMENT ATMOSPHERE MASS REMOVAL j

a. High Flow Unfiltered Vent 3 i

b. High Flow Filtered Vent 3

c. Low Flow Vent (Filtered) 2

d. Low Flow Vent (Unfiltered) 1  !

COMBUSTIBLE GAS CONTROL 6.

a. Post Accident Inerting System 3

b. Hydrogen Control by Venting 3 -

c. Pre-inerting 1

d. Ignition Systems 3 e.. Fire Suppression System Inerting 3 l

7. CONTAINMENT SPRAY SYSTEMS

a. DrywellHead Flooding 2 l

b. Containment Spray Augmentation 1 l

8. PREVENTION CONCEPTS

a. Additional Service Water Pump 2

b. Improved Operating Response I

c. DiverseInjection System 4  ;

d. Operating Experience Feedback 1 i

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, AdditionalDieselGenerator 1 58 REV O 4

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1 Table A-3 (Continued)

Modification Category

9. (Continued)

e. Increased Electrical Divisions 1

f. Improved Uninterruptable Power Supplies 1

g. AC Bus Cross-ties 1 j

h. Gas Turbine 1

i. Dedicated RHR (bunkered) Power Supply 4

10. DC POWER SUPPLIES

a. Dedicated DC Power Supply 2

b. AdditionalBatteries/ Divisions 4

c. FuelCells 4

d. DC Cross-ties I

c. Extended Station Blackout Provisions 1'

11. ATWS CAPABILITY

a. ATWS Sized Vent 2

b. Improved ATWS Capability 1

12. SEISMIC CAPABILITY

a. Increased Seismic Margins - I

b. IntegralBasemat 3

13. SYSTEM SIMPLIFICATION

a. Reactor Building Sprays 2

b. System Simplification I

c. Reduction in Reactor Bldg Flooding 1 14 CORE RETENTION DEVICES r.. Flooded Rubble Bed 2

b. Reactor Cavity Flooder 1

c. Basaltic Cements 1 59 REV O

l Table A-4 I Modifications Evaluated  ;

l

1. Accident Mang. ment la. Severe Accident EPGs/AMGs Ib. Computer Aided Instrumentation Ic. Improved Maintenance Procedures / Manuals

2. Decay Heat Removal 2a. Passive High Pressure System j 2b. Improved Depressurization 2c. Suppression PoolJockey Pump 2d. Safety Related Condensate Storage Tank

3. Containment Capability 3 a. Larger Volume Containment 3b. Increased Containment Pressure Capability 3 c. Improved Vacuum Breakers 3d. Improved Bottom Head Penetration Design

4. Containment Heat Removal 4a. Larger Volume Suppression Pool

5. Containment Atmosphere S a. Low Flow Filtered Vent Gas Removal

7. Containment Spray 7a. DrywellHead Flooding

8. Prevention Concepts 8a. Additional Service Water Pump

9. AC Power Supplies 9a. Steam Driven Turbine Generator 9b. Alternate Pump Power Source

10. DC Power Supplies 10a. Dedicated DC Power Supply

11. ATWS Capability Ila. ATWS Sized Vent

13. System Simpli6 cation 13a. Reactor Building Sprays

14. Core Retention Devices 14a. Flooded Rubble Bed 60 REV O

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Table A-5 Summary of Benefits Averted Risk PotentialImprovement Person-rem la. Severe Accident EPGs/AMGs 1.5E-2 lb. Computer Aided Instrumentation 1.0E-2 Ic. Improved Maintenance Procedures / Manuals 1.6E-2 2a. Passive High Pressure System 6.9E-2 2b. Improved Depressurization 4.2E-2 2c. Suppression Pool Jockey Pump 0.2E-2 2d. Safety Related Condensate Storage Tank 1.0E-2 3 a. Larger Volume Containment 15E-2 3b. Increased Containment Pressure Capability 16E-2 3c. Improved Vacuum Breakers 0.004E-2 3d. Improved Bottom Head Penetration Design 5.7E-2 4a. Larger Volume Suppression Pool 0.02E-2 Sa. Low Flew Filtered Vent 1.4E-2 7a. Drywell Head Flooding 6.0E-2 Sa. 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 1la. ATWS Sized Vent 3.0E-2 13a. Reactor Building Sprays 1.7E-2 14a. Flooded Rubble Bed 0.lE-2 Y

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i Table A-6 Summary of Costs Estimated Minimum Potential Improvement Cost I a. Severe Accident EPGs/AMGs S 600,000 lb. Computer Aided Instrumentation 5 599,600 1c. Improved Maintenance Procedures / Manuals S 299,000 2a. Passive High Pressure System S 1,744,000 2b. Improved Depressurization 5 598,600 ,

2c. Suppression PoolJockey Pump S 119,800 2d. Safety Related Condensate Storage Tank $ 1,000,000 3a. Larger Volume Containment S 8,000,000 3b. Increased Containment Pressure Capability 5 12,000,000 3c. Improved Vacu'im Breakers S 100,000 3d. Improved Bottom Head Penetration Design S 750,000 4a. Larger Volume Suppression Pocl S 8,000,000 2

Sa. Low Flow Filtered Vent S 3,000,000 7a. Drywell Head Flooding 5 100,000 Sa. Additional Service Water Pump S 5,999,000 9a. Steam Driven Turbine Generator S 5,994,300 9b. Alternate Pump Power Source 5 1,194,000 10a. Dedicated DC Power Supply S 3,000,000 11a. ATWS Sized Vent S 300,000 13a. ReactorBuilding Sprays $ 100,000 14a. Flooded Rubble Bed S 18,750,000 4

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Table A-7 Summary of Results Cost (K)/ Person-rem Modification Averted 7a. Drywell Head Flooding $1,667 13a. Reactor Building Sprays $5,882 1 l a. ATWS Sized Vent $10,000 3d. Improved Bottom Penetration Design $13,158 2b. Improved Depressurization S14,252 9b. Alternate Pump Power Source $17,304 1c. Improved Maintenance Procedures / Manuals 518,688 2a. Passive High Pressure System $25,275 la. Severe Accident EPGs $40,000 10a. Dedicated DC Power Supply $43,478 3 a. Larger Volume Containment $53,333 2c. Suppression Pool Jockey Pump $59,990 lb. Computer Aided Instrumentation $59,960 3 b. Increased Containment Pressure Capacity $75,000 2d. Safety Related Condensate Storage Tank S100,000 c.a Steam Driven Turbine Generator $115,275 S a. Low Flow Filtered Vent $214,286 8a. Additional Service Water Pump $374,938 3c. Improved Vacuum Breakers $2,500,000 14a. Flooded Rubble Bed $18,750,000 4a. Larger Volume Suppression Pool $40,000,000 63 REV 0