Text

Recommends That Commission Schedule No Meetings on Proposed Station Blackout Rule Until Results of Industry Findings Evaluated.Approval of Publication of Notice of Rulemaking Recommended
	ML20136G019
Person / Time
Issue date:	05/06/1985
From:	Dircks W; NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	;
References
	REF-GTECI-A-44, REF-GTECI-EL, TASK-A-44, TASK-OR, TASK-RINV, TASK-SE NUREG-1109, SECY-85-163, NUDOCS 8505310061
	Download: ML20136G019 (349)
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RULEMAKING ISSUE l (Notation Vote)

May 6, 1985 SECY-85-163 l

For: The Connissioners From: William J. Dircks Executive Director for Operations dubject: PROPOSED STATION BLACK 0UT RULE

Purpose:

To obtain Commission approval for publication of a Notice of Proposed Rulemaking on the subject of Station Blackout.

Background:

The Commission's existing regulations establish reqoirements for the design and testing of onsite and offsite electric power systems (10CFR Part 50, Appendix A, General Design Criteria 17 and 18). However, as operating experience has-accumulated, the concern has arisen that the reliability of both the offsite and onsite emergency AC power systems might be less then originally anticipated. These systems provide power for various safety systems including reactor core decay heat removal and containment heat removal which are essential for preserving the integrity of the reactor core and the containment building, respectively. The existing regulations do not require explicitly that nuclear power plants be designed to withstand the loss of all AC power for any specified period.

This issue has been studied by the staff as part of Unresolved Safety Issue (USI) A-44, " Station Blackout." Both deteministic and probabilistic analyses were performed to determine the timing and consequences of various accident sequences and to identify the dominant factors affecting the

-likelihood cf core melt accidents from station blackout.

These studies indicate that station blackout can be a significant contributor to the overall plant risk.

Contact:

Karl Kniel, NRR 492-7359

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The Commissioners Consequently, the staft is proposing that the Commission ,

amend its-regulations to require that plants be capable of withstanding a total loss of AC power for a specified

duration and to maintain reactor core cooling during that period. ,

Discussion: The staff is proposing to add a new 950.63 and to amend General Design Criterion 17 of 10CFR 50, Appendix A to '

require that all light-water-cooled nuclear power plants be capable of coping with a total loss of AC power for some

specified period of time. The time would be determined for i- each plant based on a comparison of the individual plant design with factors that have been identified as the main i contributors to risk of core melt resulting from station blackout. These factors are: (1) the redundancy of onsite emergency AC power sources, (2).the reliability of the onsite emergency AC pcwer sources, (3) the frequency of loss of offsite power and (4) the probable time needed to restore '

offsite power.

, The USI A-44 studies developed insights into the relative ;

importance of the various factors,-and those insights, along i with engineering judgement, were used to develop the proposed resolution. Irplementation of the proposed resolution wculd achieve a reasonably consistent level of

core melt frequency from station blackout' for all plants and would result in station blackout events being a relatively small contributor to total core melt frequencies identified in previous risk assessment studies.

A Notice of Proposed Rulemaking is provided as Enclosure 1,

, and a supporting Regulatory Analysis is provided as Enclosure 2. If the proposed rule were adopted, all 1 --licensees and applicants would be required: . .(1).to assess

, the capability of their plants to cope with a station

] blackout 1(i.e., determine the amount of time the plant can maintain core cooling and containment integrity with AC power unavailable); (2) to have procedures and training to cope with such an event; and (3) to make modifications, if

. necessary, to cope with a specified minimum duration station blackout selected on a plant-specific. basis. We do'not anticipate that the proposed resolution would necessitate large expenditures for plant improvements. Costs for an average plant are estimated to be about $600,000. Costs could range from an estimated $200,000 up to a maximum'of about $4 million per plant. ,

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the Comissioners ,

The staff has prepared a draft regulatory guide entitled

" Station Blackout" (Enclosure 3) which presents a method acceptable to the staff for selecting a plant-specific minimum duration for station blackout capability to comply with the proposed amendment to General Design Criterion 17.

Application of the method in this guide would result in selection of a 4- or 8-hour station blackout duration capability depending on the plant's specific design and site-related characteristics. Applicants and licensees could propose alternative methods to that specified in the draft regulatory guide for selecting the station blackout duration. The staff plans to issue the draft regulatory guide for public comment within a few days of the publication of the Notice of Proposed Rulemaking.

The staff has also prepared NUREG-1032 entitled, " Evaluation of Station Blackout Accidents at Nuclear Power Plants" which sumarizes the technical findings of USI A-44 (Enclosure 4).

This report is being printed and will be published shortly-for public coment.

The proposed resolution of USI A-44 builds on the program to improve diesel. generator reliability developed under Generic Issue B-56, " Diesel Generator Reliability." B-56 will establish a basis for maintaining and detennining the reliability of emergency diesel generators that will be used in USI A-44.

The staff intends that, if the proposed rule is published in effective form, licensees will be requested to develop, in consultation with the Office of Nuclear Reactor Regulation, proposed plant-specific schedules for implementation of modifications, if necessary, to comply with the final rule.

the proposed resolution of USI A-44 was reviewed by both the Advisory Committee on Reactor Safeguards (ACRS) and the Comittee to Review Generic Requirements (CRGR). In a letter frcm J. J. Ray to W. J. Dircks dated July 13, 1983, the ACRS supported the proposed approach for the resolution of USI A-44 but felt'that even further improvements in reactor safety may emerge from ongoing work to resolve USI A-45, " Shutdown Decay Heat Removal Requirements." .In a memo from V. Stello to W. J. Dircks dated May 8, 1984, CRGR recommended that the proposed rule, the draft regulatory guide, and NUREG-1032 documenting the staff's evaluation and proposed resolution cf USI A-44 be issued for public coment. In addition, CRGR made the following recomendation:

The Comissioners - The Statement of Considerations for the Proposed Rule should solicit comment on the feasibility and advisability of deferring implementation of the .

resolution of USI A-44 as proposed by the staff, and its potential integration with resolution of other major related and pending generic issues. Comments should be solicited for suggested means for accomplishing this integration in the most cost-beneficial way.

The optional paragraph below could be included in the Supplementary Information section of the proposed Federal Register Notice to request coments on deferring '

implementation of the resolution.of USI A-44 pending resolution of other related issues.

The Comission is seeking coments on the feasibility and advisability of deferring promulgation of the proposed rule on station blackout pending the resolution and/or integration of this Unresolved Safety Issue with cther

. related generic issues. Any coments should include suggested means for accomplishing this integration in an efficient and cost-beneficial way.

However, NRR recomends that this paragraph not be included in the Federal Register Notice. The relationship of USI A-44 to eight other related issues is discussed in Section 4b of the Regulatory Analysis. The proposed resolution of USI A-44 is compatible with the proposed or anticipated resolution of these other issues.

During the development of the proposed technical resolution of USI A-44, the. staff has had substantial interactions with various industry organizations to discuss the evolving technical studies and findings (now contained in NUREG-1032, Enclosure 4). Sumaries of the status of the program were presented at annual seminars on USIs conducted by NUS Corp. in the fall of 1982, 1983, and 1984, and technical papers were presented by the staff at several meetings of the IEEE and the ANS.

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The Comissioners In February 1982, the IEEE formed Working Group 4.3 to '

review the NRC work on station blackout, particularly the i data base being used for the technical analyses. An ANS Standards Comittee Working Group 58.12 was formed in June 1983 to develcp an industry standard that would provide a methodology for evaluating the capability of light water reactor power plants to withstand station blackout. The USI A-44 Task Managers have actively participated in meetings of both of these groups to exchange technical information and provide NRC staff perspectives on the issue. In May 1984, the ANS 58.12 Working Group issued a draft standard to ANS Nuclear Power Plant Standards Comittee (NUPPSCO) members for comment. - This draft standard is referenced in the staff's proposed Regulatory Guide (Enclosure 3). The staff understands that ANS is deferring formal balloting on the draft standard pending NRC publication of a proposed station blackout rule.

In early 1984, the Subcommittee on Probabilistic Risk

. Assessment of the AIF Committee on Reactor Licensing and Safety formed a Working Group to address the. station blackout issue. Staff representatives briefed the Working Group on the staff's approach and status in February 1984.

In a letter to the ED0 dated May 21, 1984 (Enclosure 7),

i the Comittee Chairman identified a number of technical and procedural issues that the AIF Comittee felt should be addressed before the staff submitted a proposed rule to the Comission. The staff's response, dated June 15, 1984, is provided as Enclosure 8.

In June 1984, approximately 25 utilities formed a. Nuclear Utility Group on Station Blackout (see letter to the EDO,

, Enclosura 9). Staff members met several times with representatives of this Group, and in October.1984, the Nuclear Utility Group submitted a proposed plan for NRC staff development of a resolution of USI A-44 that would be integrated with the resolution of several other issues (Enclosure 10). In a meeting with representatives of the Nuclear Utility Group on October 23, 1984, members of the staff noted that the technical studies and development of a proposed resolution performed under the Task Action Plan for the Station Blackout USI (approved in July 1980) were already substantially completed, and that the work had been performed in close coordination with work on other related issues.

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Relationships with the other issues are discussed in the '

1 The Comissioners staff's Regulatory Analysis document prepared in support of the proposed rulemaking. (The Regulatory Analysis document and the specific proposed language of the proposed rule and proposed Regulatory Guide thus far have been withheld from public disclosure as predecisional documents. They would not normally be released to the public until they are considered by the Comissicn in an open meeting.) Staff represertatives noted that the Utility Group had not had the opportunity to review the Regulatory Analysis or the language of the proposed rule, and expressed the view that further coments by the Nuclear Utility Group could appropriately be considered along with other coments received in response to a notice of proposed rulemaking.

In a letter to the Utility Group, dated November 6,1984, (Enclosure 11) the staff agreed that further dialogue between the industry and the staft would be beneficial, particularly when the staff's technical findings' report (NUREG-1032, Enclosure 4) became available for public comment. (An advance xerox copy of the report was subsequently sent to the Utility Group on January 31, 1985, and also placed in the PDR.) The November 6 letter also noted staff's intent to forward the proposed rule to the Comission in the near future, and encouraged planned efforts by the Nuclear Utility Group to organize meaningful further discussions with broad-based participation by all segments of the industry.

On February 26, 1985, the staff briefed the'ACRS Elect _rical Systems Subcommittee on the status of USI A-44 and the technical findings contained in the recently completed NUREG-1032. Presentations were also made at that meeting by EPRI and by the Nuclear Utility Group on Station Blackout.

In a letter dated March 12, 1985 (Enclosure 12), the ACRS provided favorable comments on the staff's proposed resolution.

Following the ACRS subcomittee meeting, the Nuclear Utility Group sent a letter to the OEDROGR (enclosure 13), stating their opinion that rulemaking on station blackout is un-necessary, and proposing to develop an alternative that in their view should, as a minimum, be included in the staff's proposed rulemaking package. The Nuclear Utility Group

The Commissioners requested that transmittal of the rulemaking package to the Comission be deferred for a two-month period to enable the Group to provide further input to the staff.

As scon as possible following receipt of the information promised by the Nuclear Utility Group on Station Blackout, the staff will provide the Commission with its evaluation of that information and any changes in the staff recommend-ations regarding this proposed rulemaking that may result trom that evaluation.

Recomendations: We recommend that the Commission schedule no meetings on this matter until the staff forwards its evaluation of the industry findings (about early June). That evaluation may result in changes in the staff recommendations prepared prior to receipt of the industry alternative preposals, which are that the Commission:

(1) Approve publication of the Notice of Proposed Rule-making (Enclosure 1) in the Federal Register.

(2) Note that the draft regulatory guide entitled " Station Blackout 7nclosure 3) will be issued for public comment within a few days of the publication of the Notice of Proposed Rulemaking. .

(3) Note that the Regulatory Analysis (Enclosure 2) will be published as NUREG-1109, "For Comment."

(4) Note that the Environmental Impact Statement need not be prepared since the actions proposed will not have a significant environtrental impact (see Enclosure 5). ,

(d) Note that a request will be sut,mitted to the Office of Management and Budget for review and approval of the proposed information collection requirements contained in the proposed rule.

(6) Note that the appropriate Congressional Committees will be informeo by means of a letter similar to Enclosure 6.

(7) Certify, pursuant'to requirements of the Regulatory Flexibility Act, 5 U.S.C. 605(b), that the proposed rule will not have a significant economic impact on a substantial nuraber of small entities.

The Commissioners If scheduled for a meeting, staff recommends discussion at an Open Meeting and placement of the paper in the Public Document Room.

hWillia

. Dircks o

Executive Director for Operations

Enclosures:

1. Notice of Proposed Rulemaking

2. NUREG-1109, For Comment, " Regulatory Analysis for the Resolution of Unresolved Safety Issue A-44, Station Blackout"

3. Draft Reculatory Guide, " Station Blackout"

4. NUREG-1032, For Comment, " Evaluation of Station Blackout Accidents at Nuclear Power Plants"

5. Environmental Assessment and Finding of No Significant Environmental Impact for Proposed Rulemaking to Amend 10 CFR 50 Concerning Station Blackout

6. Draft Congressional Letter

7. May 21, 1984, Letter from !!urray R. Edelman to William Dircks

8. June 15, 1984, Letter from Harold R. Denton to Murray R. Edelman

, 9. June 28, 1984, Letter from Chris H. Poindexter to William J. Dircks

10. " Integration Plan for Station Blackout (USI A-44) and Power-Related Generic Issues," Nuclear Utility Group on Station Blackcut, dated October 12, 1984

11. November 6, 1984, Letter from William J. Dircks to L. G. Kuncl and C. H. Poindexter

12. March 12,1985 Letter from David A. Ward to William J. Dircks

, 13. March 8,1985, Letter from J. Michael McGarry, III to Victor Stello, Jr.

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. Commissioners' comments should be provided directly to the Office of the Secretary by c.o.b. Wednesday, June 5, 1985.

. Commission Staff Office comments, if any, should be submitted to the Commissioners NLT Wednesday, May 22, 1985, with an information. copy to the Office of the Secretary.-~If the paper is of such a nature that it-requires additional time for analytical review and comment, the Commissioners-and the Secretariat should be apprised of when comments may be expected.

DISTRIBUTION:

Commissioners OGC OPE OI j OCA 4

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. REGIONAL OFFICES i EDO ELD 4

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9 ENCLOSURE 1 FEDERAL REGISTER NOTICE J

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[7590-01]

l Enclosure 1 NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 Station Blackout AGENCY: Nuclear Regulatory Comission.

ACTION:- Proposed rule.

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SUMMARY

The Nuclear Regulatory Comission is proposing to amend its regulations to require that light-wat,er-cooled nuclear power plants be capable of withstanding a total loss of alternating current (AC) electric power (called " station blackout") for a specified duration and to maintain reactor core cooling during that period. This proposed requirement is based on information developed under the Comission's study of dnresolved Safety Issue A-44, " Station Blackout." The proposed change is intended to provide turther assurance that a station blackout (loss of both offsite power and onsite emergency AC power systems) will not adversely affect the public health and safety.

DATE: The comment period expires [ insert a date 90 days after the publication of this Notice of Proposed Rulemaking]. Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given except as to comments received before this date.

ADDRESSES: Send comments to: The Secretary of the Comission, U.S. Nuclear Regulatory Comission, Washington, DC 20555, Attention: Docketing and Service Branch. Copies of comments received may be examined and copied for I a fee at the NRC Public. Document Room, 1717 H Street, NW, Washington, DC.

FOR FURTHER INFORMATION CONTACT: Alan Rubin, Division of Safety Technology, ,

Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Comission, _.

Washington, DC 20555, 1elephone: (301) 492-8303.

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2-SUPPLEMENTARYINFORMATIdN: The alternating current (AC) electric power for essential and nonessential service in a nuclear power plant is supplied primarily by offsite power. Redundant onsite emergency AC power systems are also provided in the event that all offsite power sources are lost. These systems provide power for various safety systems including reactor core decay heat removal and containment heat removal which are essential for preserving the integrity of the reactor core and the

. containment building, respectively. The reactor core decay heat can also be removed for a limited time period by safety systems that are independent of AC power.

The term " station blackout" means the loss of offsite AC power to the essential and nonessential electrical buses concurrent with turbine trip and the unavailability of the redundant onsite emergency AC power systems (e.g.,

as a result of units out of service for maintenance or r(pair, failure to start on demand, or failure to continue to run after start). If a station blackout persists for a sufficient time such that the capability of the AC-independent systems to remove decay heat is exceeded, core melt and containment failure could result.

The Ccmmission's existing regulations establish requirements for the design and testing of onsite and offsite electric power systems that are intended - -

to minimize the probability of losing all AC power. See General Design -

Criteria 17 and 18, 10 CFR Part 50, Appendix A. The existing regulations do not require explicitly that nuclear power plants be designed to assure that the core can be cooled and the integrity of the reactor coolant pressure boundary can be maintained for any specified period of loss of all AC power.

As operating experience has accumulated, the concern has arisen that the reliability of both the onsite and offsite emergency AC power systems might be less than originally anticipated, even for designs that meet the requirements of General Design Criteria 17 and 18,. Many operating i plants have experienced a total loss of offsite power, and more occurrences l can be expected in the future. Also, cperating experience with l

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c'nsite emergency power systems has included many instances,when diesel generators failed to start. In a few cases, there has been a ccmplete loss of both the offsite and the onsite AC power systems. During these events, AC power was restored in a short time without any serious consequences.

In 1975, the results of the Reactor Safety Study (WASH-1400) showed that

- station blackout could be an important contributor to the total risk from nuclear power plant accidents. Although this total risk'was found to be small, the relative importance of the station blackout accident was established. Subsequently, the Commission designated the issue of station blackout as an Unres'olved Safety Issue (USI); a Task Action Plan (TAP A-44) was issued in July 1980, and work was initiated to determine whether additional safety requirements were needed. Factors considered in the -

analysis of risk from station blackout included: (1) the. likelihood and duration of the loss of offsite power, (2) the reliability of the onsite AC power system, and (3) the potential for severe accident sequences after a loss of all AC power, including consideration of the capability to remove care ,

decay heat without AC power for a limited time period.

The technical findings of the staff's studies of the station blackout issue are presented in NUREG-1032, " Evaluation of Station Blackout ~ Accidents at Nuclear Power Plants, Technical Findings Related to' Unresolved Safety Issue A-44."III Additional information is provided in supporting contractor reports: NUREG/CR-3226. " Station Blackout Accident Analyses" published in

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itay 1983; NUREG/CR-2989, " Reliability of Emergency AC Power Systems at Nuclear III NUREG-1032 was issued for public comment on . Copies of this report are available for public inspection and copying for a fee at the NRC Public Document Room at 1717 H Street, NW, Washington, DC 20555.

Free single copies of NUREG-1032 may be requested by writing to the

Publication Services Section, Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555. -

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Power Plants" published in ' July 1983; and NUREG/CR-3992, " Collection and Evaluation of Complete and Partial Losses of Offsite Power at Nuclear Power Plants" published in .

(2) The major results of these studies are given below.

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Losses of offsite power can be characterized as those resulting from plant-centered faults, utility grid blackout, and severe weather-induced failures of offsite power sources. Based on operating experience, the -

frequency of total losses of offsite power in operating nuclear power plants was found to be about one per 10 site-years. The median '

restoration tirie'was about one-half hour, and 90 percent of the offsite 4

! power losses were restored in approximately 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months (NUREG/CR-3992).

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The review of a number of representative designs of onsite emergency AC power systems has indicated a variety of potentially important failure

causes. However, no single improvement was identified that could result in a significant improvement in overall diesel generator reliability. -

Data cbtained from operating experience show that the typical individual emergency diesel generator failure rate is about 2.5 x 10-2 per demand, j and that the emergency AC power system unavailability for a plant which has two emergency diesel generators, one of which is required for decay .

heat removal, is about 2 X 10-3 per demand (NUREG/CR-2989).

Given the occurrence of a station blackout, the likelihood of resultant -

, core damage or core melt is dependent on the reliability and capability of decay heat removal systems that are not dependent on AC power. If -

sufficient AC-independent capability exists, additional time will be available to restore AC power needed for long-term cooling (NUREG/CR-3226).

l I2I Copies of these documents are available for public inspection and copying for a fee at the NRC Public Document Room at 1717 H Street, NW, Washington, DC 20555. Copies may also be purchased by calling (301) 492-9530 or by writing to the Publication Services Section, Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555, or purchased from the National Technical Information Service, Department of Commerce, 5285 Port Royal Road, ~

l Springfield, VA 22161.

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It was determined by reviewing design, operational, and site-dependent factors that the expected frequency of core damage resulting from station l blackout events could be maintained near or below 10 -5 per reactor-year !

for any nuclear plant with readily achievable diesel generator 1 reliabilities, provided that the plant is designed to cope with station

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blackout for a speified duration. The duration for a specific plant is l l._-- , based on a comparison of the plant's characteristics to those factors .

. that have been identified as the main contributors to risk from station j blackout (NUREG-1032). 1

) As a result of the station blackout studies, improved guidance will be j provided to licensees regarding maintaining minimimum emergency diesel I generator reliability to minimize the probability of losing all AC power.

In addition, the Conunission is proposing to amend its regulations by adding a new $50.63 and by adding a new final paragraph to General Design Criterion i 17, Appendix A of 10 CFR Part 50, to require that all nuclear power plants be

[ capable of coping with a station blackout for some specified period of time. ,

The period of time for a specific plant would be determined based on the existing capability of the plant as well as a comparison of the individual plant design with factors that have been identified as the main contributors to risk of core melt resulting from station blackout.

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.These factors, which vary significantly from plant to plant because of i considerable differences in design of plant electric power systems as well as d site-specific considerations, include: (1) redundancy of onsite emergency AC l

l power sources (i.e., number.of sources minus the number needed for decay heat !

j removal), (2) reliability of onsite emergency AC power sources (usually

! diesel generators), (3) frequency of loss of offsite power, and (4) probable (

time to restore offsite power. The frequency of loss of, and time to restore j offsite power are related to grid and switchyard reliabilities, historical !

! weather data for severe storms, and the availability of nearby alternate l power sources (e.g., gas turbines). Experience has shown,that long duration '!

l offsite power outages are caused primarily by severe storms (hurricanes, ice.

l snow,etc.).

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The objective of'the proposed rule is to reduce th'e risk of severe accidents resulting from station blackout by maintaining highly reliable AC electric power systems and, as additional defense-in-depth, assuring that plants can cope with a station blackout for some period of time. If the proposed rule is adopted, all licensees and applicants would be required to assess the capability of their plants to cope with a station blackout (i.e., determine

- the amount of time'the plant can maintain core cooling and containment

integrity with AC power unavailable), and to have procedures and training to cope with such an event. Plants would be required to be able to cope with a

specified minimum duration station blackout selected on a plant-specific basis.' '

On the basis of station blackout studies conducted fcr USI A-44, and presented in the reports referenced above, the NRC staff has developed a -

draft regulatory guide entitled " Station Blackout,"(3) which presents guidance on ,(1) maintaining a high level of reliability for emergency diesel generators, (2) developing procedures and training to restore offsite and i onsite emergency AC power should either one or both become unavailable, and (3) selecting a plant-specific minimum duration for station blackout capability to comply with the proposed amendment to General Design Criterion

17. Application of the methods in this guide would result in selection of a 4-hour or 8-hour station blackout duration, depending on the specific plant design and site-related characteristics. However, applicants and licensees could propose alternative methods to that specified in the regulatory guide in order to justify other minimum durations for station blackout capability.

(3)A notice of availability and request for comments on the draft regulatory guide will be published within a few days of this Notice of Proposed Rulemaking. Copies of the draft regulatory guide are available for public inspection and copying for a fee at the NRC Public Document Room at 1717 H Street, NW, Washington, DC 20555, and will be distributed to those on the automatic distribution list for draft regulatory guides. Free single copies of the draft regulatory guide may be obtained by writing to the U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention:

Director, Division of Technical Information and Document Control.

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If the proposed rule and regulatory guide are issued, those plants with an already low risk from station blackout would be required to withstand a station blackout for a relatively short period of time and probably would need few, if any, modifications as a result of the rule. Plants with currently higher risk from station blackout would be required to withstand somewhat longer duration blackouts. Depending on their existing capability, these pl. ants might also need to make modifications (such as increasing station battery capacity or condensate storage tank capacity) in order to cope with the longer station blackout duration. The proposed rule would require licensees to develop, in consultation with the Office of Nuclear .

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'ReactorRegu1'ation,proposedplant-specificschedulesfokim'plementationof any needed modifications.

FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT: AVAILABILITY -

The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment, and therefore an environmental impact statement is not required. There would not be any adverse environmental impacts as a result of the proposed rule for the tollowing reasons: (1) there would be no additional radiological exposure to the general public or plant employees, and (2) plant shutdown is not ---

required so there would be no additional environmental . impacts as a result of the need for replacement power. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 1717 H Street NW, Washington, DC. Single copies of the environmental assessment and the finding of no significant impact are available from Mr. Karl Kniel, Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone: (301) 492-7359.

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PAPERWORK REDUCTION ACT STATEMENT This proposed rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.).

This rule has been submitted to the Office of Management and Budget for review and approval of the paperwork requirements.

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REGULATORY ANALYSIS The Comission has prepared a regulatory analysis for this regulation. The

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analysis examines the costs and benefits of'the rul'e as considered by the

Comission. A copy of the regulatory analysis, NUREG-1109, For Comment,

" Regulatory Analysis for the Resolution of Unresolved Safety Issue A-44, Station, Blackout," is available for inspection and copying for a fee at the NRC Puelic Document Room, 1717 H Street, NW, Washington, DC 20555. Free single copies of NUREG-1109 may be obtained by writing to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Comission, Washington, DC 20b55.

The Ccmission requests public comment on the regulatory analysis. Comments on the draft analysis may be submitted to the NRC as indicated under the ADDRESSES heading.

REGULATORY FLEXIBILITY CERTIFICATION.

In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b),

the Comission hereby certifies that this proposed rule, if promulgated, will not have a significant economic impact on a substantial number of small entities. This proposed rule specifies that nuclear power plants be able to withstand a total loss of AC power for a specified time duration and maintain reactor core cooling during that period. These facilities are licensed under the provisions of 10 CFR 50.21(b) and 10 CFR 50.22. The companies that own these facilities do not fall within the secpe of "small entities" as set forth in the Regulatory Flexibility Act or the small business size standards set forth in regulations issued by the Small ,

Business Administration in 13 CFR Part 121. _

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LIST OF SUBJECTS IN 10 CFR PART 50 Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, notice is hereby given that adoption of the'following amendmen'ts to 10 CFR Part 50 is contemplated.

PART 50 ' DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES

1. The authority citation for Part 50 continues to read as follows:

AUTHORITY: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2133, 2134, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended (42 U.S.C. 5841, 5842, 5846), unless otherwise noted.

Section 50.7 also issued under Pub. L.95-601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851). Sections 50.57(d), 50.58, 50.91, and 50.92 also issued under

! Pub. L.97-415, 96 Stat. 2071, 2073 (42 U.S.C. 2133, 2239). Section 50.78-also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C.

2234). Sections 50.100-50.102 also issued under sec. 186, 68 Stat. 955 (42 U.S.C. 2236).

For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C. 2273),

5 50.10(a), (b), and (c), 50.44, 50.46, 50.48, 50.54, and 50.80(a) are issued under sec. 161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b));

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10 50.10(b) and (c) and 50.54 are issued under sec. 1611, 68 Stat. 949, as amended (42 U.S.C. 2201(1)); and 50.55(e),50.59(b),50.70,50.71, 50.72, 50.73, and 50.78 are issued under sec. 1610, 68 Stat. 950, as amended (42 U.S.C. 2201(o)).

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2. In 50.2, a definition of " station blackout" is added as a new paragraph (y) to read as follows:

550.2 Definitions.

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(y) " Station blackout" means the complete loss of alternating current (AC) electric power to the essential and nonessential switchgear buses in a nuclear power plant (i.e., loss of the offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency AC power system).

3. A new 50.63 is added to read as follows:

s50.63 Loss of All Alternating Current Power.

(a) Requirements. Each light-water-cooled nuclear power plant licensed -

to operate must be able to withstand and recover from a station blackout as defined in 50.2 for a specified duration in accordance with the requirements in paragraph (e) of General Design Criterion 17 of Appendix A of this part.

(b) Limitation of Scope. Paragraphs (c) and (d) of this section do not apply to those plants licensed to operate prior to [ insert the effective date of this amendment] if the capability to withstand station blackout was considered in the operating license proceeding and a specified duration was accepted as the licensing basis for the facility.

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(c) Implementation - Determination of Station Blackout Duration.

(1) For each light-water-cooled nuclear power plant licensed to operate on or before [ insert the effective date of this amendment], the licensee shall submit to the Director of the Office of Nuclear Reactor Regulation by (insert a date 270 days after after the effective date of this amendment]:

(i) a determination of the maximum duration for which the plant as currently designed is able to maintain core cooling and containment integrity in the event of a station blackout as defined in 550.2(y);

(ii) a description of the procedures that have been'est'ablished for stationblackouteventsfortheduration/tterminedinparagraph(c)(1)(1) of this section and for recovery therefrom; (iii) an identification of the factor (s) that limit the capability of the plant to cope with a station blackout for a longer time than that -

determinedinparagraph(c)(1)(1)ofthissection; (iv) a proposed station blackout duration to be used in determining compliance with paragraph (e) of General Design Criterion 17 of Appendix A of this part, including a justification for the selection based on: (1) the redundancy of the onsite emergency AC power sources, (2) the reliability of the onsite emergency AC power sources, (3) the expected frequency of loss _

.of offsite power, and (4) the probable time needed to restore offsite power; and (v) an identification of the factors, if any, that limit the capability of the plant to meet the requirements of Criterion 17 for the specified station blackout duration proposed in the response to paragraph (c)(1)(iv)ofthissection.

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(2) After consideration of the information submitted in accordance with paragraph (c)(1) of this section, the Commission will notify the licensee of its determination of the specified station blackout duration to be used in determining compliance with General Design Criterion 17 of Appendix A of this part.

._ _._ 1d )._ Implementation Schedule for Implementing Equipment Modifications._._ ;_ ___

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(1) For each light-water-cooled nuclear power plant licensed to operate -

on or before [ insert the effective date of this amendment), the licensee shall, within 180 days of the notification provided in accordance with paragraph (c)(2) of this section, submit to the Director of the Office of Nuclear Reactor Regulation a schedule for implementing any equipment and procedure mcdifications necessary to meet the requirements of General Desig~n Criterion 17 of Appendix A of this part. This submittal must include an explanation of the schedule and a justification if the schedule does not I provide for completion of t,he modifications within two years of the notification provided in accordance with paragraph (c)(2) of this section.

(2) The licensee and the NRC staff shall mutually agree upon a final schedule for implementing modifications necessary to comply with the requirements of Criterton 17. -

4. In Appendix A, General Design Criterion 17 is revised to read as follows:

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APPENDIX A General Design Criteria for Nuclear Power Plants l 3. ._;_ I. Overall Requirements * * * -. - -

. Criterion 17-Electric Power Systems. (a) An onsite electric power . .

system and an offsite electric power system shall be provided to permit functioning of s'tructures, systems,'and components important to safety. The safety function for each ' system (assuming the other system.is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is. cooled and containment integrity and other vital functions are maintained in the event of postulated -

accidents.

(b)- The onsite electric power supplies, inc.luding the batteries, and the onsite electric distribution system, shall have sufficient independence, _.

redundancy, and testability to parform their safety functions assuming a single failure.

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(c) Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to both circuits is acceptable.- Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating current' power supplies and the other offsite electric <

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power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded.

One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained.

(d) Provisions shall be included to minimize the probability of losing

. electric power from any of the remaining supplies as a' result of, or -

coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power 'supp1'iesi (e) The reactor core and associated coolant, control, and protection systems, including the station batteries, shall provide sufficient capacity

___ and capability to assure that the erre is cooled and containment integrity is _ _ _ _ _

maintained in the event of a station blackout (as defined in 450.2(y)) for a specified duration. The following factors shall be considered in specifying _ . _

the station blackout duration: (1) the redundancy of the onsite emergency AC power sources, (2) the reliability of the onsite emergency AC power sources, (3) the expected frequency of loss of offsite power, and ,_ . _ _ _ _

(4) the probable time needed to restore offsitie power.__ __

Dated at Washington, DC, this day of 1984.

For the Nuclear Regulatory Commission Samuel J. Chilk, Secretary of the Commission.

Underlined text indicates proposed additional paragraph to GDC 17.

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ENCLOSURE 2 REGULATORY ANALYSIS

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ENCLOSURE 2 NUREG-1109 For Comment Regulatory Analysis for the Resolution of Unresolved Safety Issue A-44, Station Blackout

1. Statement of the Problem .

The term " station blackout" refers to the complete loss of alternating current (AC) electric power to the essential and nonessential switchgear buses in a nuclear power plant. Station blackout therefore involves the loss ,

of offsite pcwer concurrent with turbine trip and the unavailability of the onsite emergency AC power system. Becruse many safety systems required for reactor core decay heat. removal and containment heat removal are dependent on AC power, the consequences of station blackout could be severe.

1he concern abcut station blackout' arose because of the acce.nulated experience regarding the reliability of AC power supplies. There have been numerous instances of emergency diesel generators foiling to start and run in response to tests conducted at operating plants. In addition, a number of cperating plants have experienced a total loss of'offsite electric power, and more occurrences are expected in the future. In almost every one of these loss-of-offsite-power events, the onsite emergency AC power supplies were available imediately to supply the power needed by vital safety equipment. However, in some instances, one of the redundant emergency power supplies has been unavailable. In a few cases there has been a complete loss of AC power, but during these events, AC power was restored in a short time without any serious consequences.

The results cf the Reactor Safety Study (Reference 1) showed that for one of the two plants evaluated, a station blackout accident could be an important contributor to the total risk from nuclear power plant accidents. Although this total risk was found to be small, the relative importance of the station blackout accident was established. This finding and the accumulated diesel generator failure experience increased the concern about station blackout.

.The issue of station blackout involves the likelihood and duration of the loss of offsite power, the redundancy and reliability of onsite emergency AC power systems, and the potential for severe accident sequences after a loss of all AC power. These topics were investigated under Unresolved Safety Issue Task Action Plan A-44, and the technical findings are reported in detail in References 2 - 5. In addition to identifying important factors and

. sequences that could potentially lead to station blackout, the results

. indicated that estimated core melt

frequencies from station blackout vary -

significantly for different plants but could be on the order of 10'4 per reactor-year for some plants. In order to reduce this risk, action should be taken to resolve the safety concern stemming' fr'om station blackout. The issue is of concern for both PWRs and BURS.

There is presently no requirement for plants to be able t,o cope with a st< tion blackout. Existing requirements for offsite and onsite AC power systems are contained in General Design Criterion 17, " Electric Power Systems," and discussed 1n Standard Review Plan (SRP) Sections 8.2, "Offsite Power Systems," and 8.3.1, "AC Power Systems (Onsite)." Testing of emergency diesel generators is discussed in Regulatory Guide 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants." Separation and independence of electric power systems are discussed in Regulatory Guide 1.6, " Independence Between Redundant Standby

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(Onsite) Pcwer Sources and Between Their Distribution Systems," and Regu-latory Guide 1.75, " Physical Independence of Electric Systems." SRP Section - -

Analysis has shown that for postulated station blackout events, the difference between the estimated frequency of core damage and core melt is small because of the relatively low probability of recovering AC power and terminating an accident sequence at a time between initial core damage and full core melt (Ref. 5).

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9.5 discusses design and maintenance provisions for the onsite emergency diesels. All of the above licensing requirements and guidance are directed at providing reliable offsite and onsite AC power.

Experience has shown that there are practical limits in assuring the reliability of offsite and onsite emergency AC power systems. Analysis

.shows that the ability to withstand a total loss of AC power until either

-~ offsite or onsite emergency AC power can be restored can significantly reduce core melt frequency.

2. ' Objectives The general cbjective of the proposed requirements is to reduce the risk of severe accidents associated with station blackout by makipg station blackout a relatively small centributor to total core melt frequency. Specific actions called for in the proposed resolution include: (1) maintaining highly ,

reliable AC electric power systems; (2) deve1cping procedures and training to restore offsite and onsite emergency AC power should either one or both become unavailable; and, as additional defense-in-depth, (3) assuring that plants can cope with a station blackout for some period of time based on the prob-ability of occurrence of a station blackout at the site as well as the AC pcwer restoration capability for that site.

3. Alternatives The following approaches were considered as alternatives to resolve USI A-44,

a. Proposed Actions To achieve the objectives stated in Section 2 above, the proposed resolution of USI A-44 calls for specific guidance relating to offsite and

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4 onsite emergency AC p6wer systems as well as a requirement for plants to be able to cope with a station blackout. A summary of the proposed recommendations to resolve this issue is presented below.

(1) The reliability of the onsite emergency AC power sources should be maintained at or above specified acceptable reliability levels.

- - = (ii) Procedures and training should be developed to restore emergency '-

r-AC power and offsite power using nearby power sources in the event that the emergency AC power system and the normal offsite power systems are unavailable. '

(iii) Each nuclear power plant should be evaluated to determine its capability to withstand and recover from a station blackout. This evaluation should include: -

Yerifying adequacy of station ba,ttery power, condensate storage -

tank capacity, and plant / instrument air for the station blackout duration.

Yerifying adequate reactor coolant pump seal integrity for the station blackout duration by demonstrating, via experiment and/or .

analysis, that seal leakage due to lack of seal cooling will not result in a sufficient primary system coolant inventory reduction -

to lose the ability to cool the core during the station blackout duration.

Verifying operability of equipment needed to cperate during a station blackout for environmental conditions associated with total loss of AC pcwer (i.e., loss of heating, ventilation and air condicioning).

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5-(iv) Each nuclear power plant should be able to withstand and recover from a station blackout lasting a specified minimum duration. A draft regulatory guide entitled " Station Blackout" (Ref. 6) provides a method to determine an acceptabit plant-specific station blackoutdurationbasedona.comparisonoft)eplant's -

characteristics to those factors that have been identified as the main contributors to risk from station blackout. These factors '

include: (1) the redundancy of onsite emergency AC power sources - -

(number of souces available for decay heat removal minus the - -

number needed for decay heat removal), (2) the reliability of onsite emergency AC power sources (usually diesel generators)', (3) '

the frequency of loss of offsite power, and (4) the probable time to restore offsite power. The frequency and duration of loss of offsite power are related to grid and switchyar,d reliabilities,

( historical weather dats for severe storms, and the availability of nearby alternate power sources (e.g., gas turbines). Reference 5 concluded that long-duration offsite power outages are caused primarily by severe storms (hurricanes, ice, snow, etc.)-

(v) If the plant's station blackout capability, as determined in (iii),

is significantly less than the recommended plant-specific station -

blackout duration in (iv), then modifications to the plant may be 3 necessary to extend the time the plant is able to cope with a sta- - - -

tion blackout. The draft regulatory guide identifies specific: - - -

factors to be considered if such modifications are necessary.

(vi) Each nuclear power plant should have procedures and training to -

cope with a station blackout and to restore normal long-term decay heat removal once AC power is restored. ,

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Sincethereispres'entlynorequiNmen'tforp1antstobe'abletowithstanda l loss of both the offsite and onsite emergency AC power systems, the proposed resolution calls for rulemaking to require all plants to be able to cope with a station blackout for a specified duration. The draft regulatory 4

guide describes a method acceptab'le to the NRC staff for complying with the proposed rule, and, in addition, specifies guidance on providing reliable AC

.- ~ - - electric power supplies. Plants with an already low risk.from station , -

blackout would be required to withstand a station blackout for a relatively -

'short period of time. These plants probably would need few, if any, modifi cations as a result of the proposed rule. Plants with currently higher risk

- front station blackdut would' be required to withstand somewhat longer duration i

l blackouts, and, depending on their existing capability, may need to make i modifications (such as increasing station battery capacity or condensate storage tank capacity) in order to meet the longer station blackopt duration. These modifications are judged by the staff to be cost-effective in terms of reducing risk to the public.

2 The method to determine an acceptable station blackout' duration capability. -

as presented in the draft regulatory guide, is sunnarized below. Appli-cations of this guide would result in a 4-hour or 8-hour station blackout duration (see Table 1) depending on the plant's specific design and site-

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related characteristics. Licensees may propose durations different from:-- ::. l those specified in Table 1. The basis for alternate durations would be -

predicted on plant-specific factors relating to the reliability of AC power- :--

systems such as those discussed in Reference 5. ,

Tables 2 and 3 provide descriptions of emergency AC power configuration groups and offsite power design characteristic groups, respectively, used in Table 1. Table 2 describes plants according to different levels of redundancy of the onsite' emergency AC power system. Table 3 provides a method to determine characteristic loss of offsite power frequency and 4 duration groups as a function of expected frequency of severe weather events, switchyard designs, number of oftsite power circuits, and ability to restore j offsite power.- After identifying the appropriate groups from Tables 2 and 3,

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Table 1. Acceptable Station Blackout Duration Capability (hours)I Emergency AC Power Configuration Group 2 A B C

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3 Maximum EDG Failure Rate Per Demand Offsite Power Design Characteristics 4 0.05 0.025 0.05 0.025 l

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'sssis,A,,,,,,f k@ w'"- "+8 I Variations from these times will be considered by the staff if justification, including a cost-benefit analysis, is provided by the licensee. The methodology and sensitivity studies presented in NUREG-1032 (Ref. 5) are acceptable for use in this justification.

2 See Table 2 to determine emergency AC power configuration group.

3 Number of emergency diesel generator (EDG) failures in the last 100 valid demands divided by 100.

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4 See Table 3 to determine categories P1, P2, and P3.

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. . l Table 2. Emergency AC (EAC) Power Configuration Groups Emergency AC power Number of EAC power Number of EAC power sources required l to operate AC powered decay heat configuration group sources available removal systems 2 1.

P 3 1 A

4 1 5 2 3 1' B 2 4 2 4 1 C 2 .

3 2 4 3 5 3 IIflany of the EAC power sources are shared between units at a multi-unit site, this is the total number of shared and dedicated sources for those - _ -

units at the site. . ,

2This number is based on all the AC loads required to remove decay heat ~

(including AC-powered decay heat removal systems) to achieve and maintain' hot shutdown at all units at the site with offsite power unavailable. - .

For nonshared EAC power sources. --

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! For shared EAC. power sources in which each diesel generator is capable of l

providing AC power to both units at a site concurrently.

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l Table 3. Offsite Power Design Characteristic Groups Group Offsite Power Design Characteristic Sites which have either one of the following offsite power designs: .

- Sites with all offsite power sources connected to the plant through two or more switchyards or separate incoming transmission lines, with at least one of the AC sources electrically independent of the others, or P1 - Sites with all offsite power sources connected to the plant

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through one switchyard or through two or more switchyards = =

that are electrically connected, and if the normal AC power source is lost, there is an automatic transfer to an alternate offsite power source. If this source fails also, there is one or more automatic or manual transfers of power to another source of offsite power.

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Sites with frequency of loss of offsite power due to extremely severe weather less than 1 per 350 site years.*

and ,,

Sites which have one or both of the following characteristics:

- The capability and procedures to recover offsite (non-emergency) AC power to the site within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following a-loss of offsite power due to severe weather,-or- -- - - - - -

- A frequency of loss of offsite power due to extremely severe weather less than 1 per 100 site years.** ~~

Sites with frequency of loss of offsite power due to extremely '

severe weather less than 1 per 100 site years *, and any one or more of the following three characteristics: -

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1. Sites with frequency of loss of offsite power'due to - N extremely severe weather greater than 1 per 350 site-years *,

or P2 2. Sites with all offsite power sources connected to the -

plant through one switchyard or through two or more switch-yards that are electrically connected, and if the normal source of AC power is lost, there are no automatic transfers and one or more manual transfers (or alternately, there is one automatic transfer but no manual transfers) to preferred or alternate offsite power sources. I or 3.Siteswithfrequencyoflossofoffsitepowerdueto severe weather greater than 1 per 100 site years **; without the capability and procedures to recover offsite (non-emergency) AC power to the site within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following a loss of offsite power due to severe weather.

P3 Sites with frequency of loss of offsite power due to extremely .

severe weather greater than 1 per 100 site years *. ,

Footnotes for Table 3

The estimated frequency of loss of offsite power due to extremely severe weather is determined by the annual expectation of storms at the site with wind velocities greater than or equal to 125 mph.

- The estimated frequency of loss of offsite power due'to severe weather, f, is determined by the following equation: -

f = (1.8 x 10-4) h1 + (27) h2 + (2.6 x 10-2) 3 h where hy = annual expectation of snowfall for the site, in inches, h2 = annual expectation of tornadoes per square mile at the site, h3 = annual expectation of storms at the site.with wind velocities between 75 and 124 mph, and The annual expectation of snowfall, tornadoes and storms are obtained from National Weather Service data from the weather station nearest to the -

plant or an interpolation, if appropriate, between nearby weather stations.

The basis for the empirical equation for the frequency of loss of offsite power due to severe weather, f, is given in Reference 5, Appendix A.

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4 and the reliability level of the onsite emergency AC po'erw sources, Table 1 can be used to determine the station blackout duration guidelines for each plant (4 or 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ). The reliable operation of the onsite emergency AC power sources should be assured by a reliability program designed to monitor and maintain reliability over time at a specified acceptable level and to improve the reliability if that level is not achieved.

As an example of an application of this method, consider a nuclear power plant that has two diesel generators, one of which is required for AC power for decay heat removal systems; an estimated frequency of loss of offsite power'due to extremely severe weather of less than 0.01 per year; one switch-yard; and one alternate offsite power circuit in addition to the normally energized offsite circuit to the IE buses. From Table 2, this plant is in Emergency AC Power Configuration Group B, and from Table 3, in Offsite Power ,

Design Characteristic Group P2. As indicated on Table 1, if the failure rate of each emergency diesel generator is maintained at 0.025 or less, this plant should have the capability to withstand and recover from a station blackout

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lasting 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or more. If the failure rate of each emergency diesel ,

generator were between 0.025 and 0.050, the acceptable station blackout duration would increase to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . If the emergency diesel generator failure rate were greater than 0.05, then steps should be _taken to improve the diesel generator reliability in accordance with the reliability program discussed above.. ,

b. Alternative (ii)

Treat plants uniformly by requiring all plants to be able to cope with the same station blackout duration.

c. Alternative (iii)

Require plants with the highest potential risk from station blackout to add either an additional emergency diesel generator or another AC-independent decay heat removal system.

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d. Alternative (iv)

Take no action.

4. Consequences

a. Costs and Benefits of Alternatives -

Alternative (i) - Proposed Action -

The benefit from implementing the pro ~ posed rule'and the guidance of the draft regulatory guide (Reference 6) is a reduction in the frequency of core melt per reactor-year due to station blackout and the associated ,

risk of offsite radioactive releases. The costs are primarily for industry to develop procedues, to improve diesel generator reliability

. if the reliability falls below certain levels, and to retrofit plants with additional components or system, as necessary, to meet the proposed requirements.

(a) Value -

Risk Reduction Estimates ,

In order to estimate the change in expected risk that the proposed - - -

resolution could effect, both the postulated radioactive exposure :-- -

(in person-rem) that would result in the event of an accident and the -

reduction in frequency of core melt must be estimated. A simplified method to estimate public dose for value-impact analysis could utilize an " average" plant to estimate the consequences of station blackout and subsequent core melt for all plants. For example, the total exposure within a 50 mile radius of a " typical" site is around 5 million person-rem (Reference 5, Section 7). However, use of a single value does not account for different offsite consequences associated with l different reactor sizes and population density differences among sites.

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'Bec5use of the differdnces between sites, as well as plant designs, it was not feasible to select a " typical" plant for analysis, using the value and impacts for that plant and then multiplying by the total number of plants to obtain an overall value-impact ratio. Instead, the

' approach taken was to treat a number of plants having similar design characteristics together, thereby reflecting more accurately the total values and impacts, yet not having to perform a plant-specific value-impact analysis for all plants. The method described below was used to calculate estimated offsite consequences for use in this value-impact -

analysis. Results indicate that consequences range from 0.5 to 7 million person-rem, witfi an average of about 2 million person-rem'per plant.

Estimates of offsite consequences of potential accidgnts at nuclear power plants are presented in Reference 7. This report includes results of calculations for 91 sites in the United States which had reactors with operating licenses or construction permits. Actual population distributions around the sites were taken into account in calculating estimated total population dose (person-rem) for various fission product releases. Results presented for each site include a scaling-factor to account for different reactor power levels at the various sites.

The scaled results from Reference 7 for release category.SST1-(siting --

source term) were used to develop estimates of site-specific- -

consequences for station blackout events.* However, these results were not used directly in the value-impact anlaysis for the following-reasons. First, the consequences in Reference 7 include the entire population around the plant (i.e., an infinite radius), whereas Enclosure 1 of Reference 8 specifies using a 50 mile radius around the i

Five release categories have been defined by NRC, denoted as SST1-SST5, to i represent a spectrum of five accident groups. .Each category represents a different degree of core degradation and failure of containment safety teatures. Group 1, SST1, is the most severe and involves a loss of all installed safety features and direct breach of containment.

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plant to calculate risk reduction estimates. Second, SST1 over-

, estimates the fission product release for station blackout events.

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Scaling factors comparing offsite exposure within a 50 mile radius of a plant to that for an infinite radius are included in Table 3 of Reference 9. Depending on the containment type, the total person-rem

- : _ . - within a 50 mile radius is between 1/3 and 1/5 the person-rem for an - -

infinite radius. These fact 6rs were used to scale the site-specific ,

results from Reference 7. -

In addition, comparing ~the fission product releases for station b1'ackout'

, from the " typical" plant in Reference 5 with SST1 releases for the equivalent plant in terms of size and population distribution in Reference 7, indicates that the SST1, offsite person-rem consequences are

^ - about a factor of three greater than the consequences from station

{' - blackout events. The main reason for this is because SST,1 is an estimated upper bound on risk assuming prompt containment -failure,'-

whereas given a core melt resulting from station blackout, containment failure would be delayed for a number of hours. Therefore, the site-specific SST1 consequences from Reference 7 were reduced by a factor of three to estimate the person-rem per accident for use in this value- - .

impact analysis.

To clarify the method discussed above, an example calculation is given -- -

- : .: . =

t for an 845 MWe PWR, Calvert Cliffs. From Appendix A of Reference 7, the 7 -

- scaled result for offsite effects is 3.61 x 10 person-rem. This number is multiplied by 1/5 to account for the 50 mile radius (Reference 9)andalsoby1/3toaccountforthesmallerlreleasesforstation l blackout events compared to SST1 releases. The resulting offsite exposure from a station blackout event and subsequent core melt within a 50 mile radius of the plant is estimated to be about 21 million person-rem.

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The reduction in frequency of core melt resulting from the proposed l resolution was estimated for groups of plants' having similar ,

characteristics rather than for each plant. Table 1 provides the ~

framework for selecting.the plant groupings. plant- and site-specific characteristics for a total of 67 reactors, which represent most of the f currently operating nuclear power plants, were categorized according to i< -

. the factors in Table 1. Tabl,e 4 presents an estimate, based"on~ -

- - .. inforination available to NRC, of the number of reactors-having these. . .

specific characteristics. -

The estimate'd core m'elt frequency reductions for each group were based on results of analyses of core melt frequency as a function of a plant's

(

ability to cope with a station blackout (Reference 5). Core melt frequency reductions per reactor-year for each group, of plants were ,

based on plants meeting the 4- or 8-hour acceptable station blackout duration in the proposed resolution. The mean reduction in core melt frequency for all plants is approximately 3 x 10-5 per; reactor-year.

This was based on the assumption that 50 percent of the plants have .

diesel generators with failure rates of 0.025 failures per demand or -

less. It was also assumed that all plants as currently designed can

. cope with a station blackout for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and, with proper procedures- -

. and training, plants could cope with a 4-hour station blackout without having to make major modifications.

Examples of the reduction in frequency of core melt per reactor-year for

three cases are presented in Table 5. The first example is typical of a-plant with one redundant' emergency AC power system (e.g., one out of ~ two diesel generators required for emergency AC power), a failure rate of ;

0.025 failures per. demand for each diesel generator, and a plant located '

in an area with average loss of offsite power duration and frequency.

The second case, which is typical of a plant with less desirable .

characteristics from a station blackout' perspective (e.g., minimum

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Table 4. Estimated Number of Reactors Having Similar Characteristics Emergency AC Power Configuration Group I .

Group A B C Total Estimated Number 11 42 14 57 of Reactors .

. Offsite Power Design Characteristics -2 Characteristic P1 P2 P3 Total Estimated Number 24 16 27 67 of Reactors I

See Table 2 for definition of emergency AC power configuration group.

2 See Table 3 to determine offsite power design characteristics.

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I Table 5. Examples of Reduction in Frequency of Core Melt Per Reactor-Year Estimated Core Damage Estimated Reduction in Plant Frequency Per Core Damage Frequency Characteristics Reactor-Year Per Reactor-Year

-5 Plants with one out of two 2.1 x 10 with 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.3 x 10 -5 emergency diesel generators station blackout capability (EDG); EDG failure rate of f 0.025 failures per demand; 0.8 x 10 -5 with 4* hour -

j; and loss of offsite power station blackout capability design characteristic group P2.

Plants with two out of three EDG; 15 x 10

-5 with 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 14 x 10-5 ,

EDG failure rate of 0.05 failures station blackout capability -

per demand; and loss of offsite N power design characteristic group P3. '

1 x 10 ~5 with 8* hour station blackout capability Plants with one out of three EDG; 0.2 x 10 -5 with 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months -

EDG failure rate of 0.025 failures station b.lackout capability 0.1 x 10 -5 ,

per demand; and loss of offsite power design characteristici group Pl. '

! 0.1:x 10'5 with 4* hour l station blackout capability 1

These times are the acceptable station blackout durations from Table 1 for these example cases.

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. o redundant emergency AC power system, below average diesel generator reliability, relatively more frequent and longer duration loss of offsite power history), has an order of magnitude larger reduction in frequency of core melt. The third case is for plants having more favorable characteristics than the base case and therefore a correspondingly lower 2

reduction in core melt frequency. ,

The estimated risk reduction for each group of plants was calculated by -

multiplying the reduction in core melt frequency per reactor-year for that group by the following three factors: (1) the number of reactors in the group, (2) the average remaining life of the plant, and (3) the estimated average public dose (in person-rem) that would result in the event of an accident. The person-rem reduction for each group of plants was then added together to calculate the total estimated risk reduction

. associated with the proposed resolution of USI A-44. Assuming an average 25 year remaining plant life, the estimated total dose reduction is 80,000 person-rem.

(b) Impacts -

Cost Estimates The cost for licensees to comply with the proposed requirements will vary depending on the existing capability of each plant to cope with a

~~

= station blackout as well as the. plant-specific station blackout duration -

a

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as determined from Table 1. It is anticipated that almost all plants snould be able to meet a 4-hour duration guideline without major plant modifications. Some plants, in addition to being able to withstand a 4-hour station blackout, may be capable of coping for longer periods without major modifications. To meet an 8-hour guideline, some plants may have to increase the capacity of one or more of the following systems: station batteries, condensate storage tank, and instrument or i

compressed air. Shedding nonessential loads from the station batteries f could be considered as an option to extend the time until battery depletion. Corresponding procedures for load shedding would need to be incorporated in the plant-specific technical guidelines and emergency operating procedures for station blackout.

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Those plants that can not verify adequate reactor' coolant p' ump seal' integrity for the station blackout duration may have to add an AC-independent charging pump for reactor coolant pump seal cooling to maintain seal integrity and adequate reactor coolant inventory.

Table 6 presents cost estimates of possible hardware modifications and procedures that could result from implementation of the recommendations proposed to resolve the station blackout issue. Since the duration- .

guidelines are based on plant-specific features, and the existing capability of systems and components needed during a station blackout vary frein plant to < plant, the modifications id Table 6 may be needed at seme, but not all nuclear power plants. For each proposed modification the table identifies an estimated range of costs per plant, the

! estimated number of plants needing that modif'ication, and the estimated total cost. The estimated total cost for industry to comply with the proposed resolution of USI A-44 is about $40 million. The average cost per reactor would be around $600,000 with a range of from $200,000, if only a station blackout assessment and procedures and training'are necessary, up to a maximum of about $4 million if modifications 1 through 4 are needed including requalification of a diesel generator.

Including averted plant damage costs can significantly affect the overall cost-benefit evaluation. The effect of the proposed action on -

averting plant damage and cleanup costs has been estimated by -

multiplying the reduction in accident frequency by the discounted onsite property costs. The following equations from Reference 12 were used to make this calculation:

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I lable 6. Estimated Costs for Industry to Comply with the Proposed Resolution of USI A-44 III Estimatyp)CostPer Estimated Total Cost ($1,000)

Estimated Number of Reactor ($1,000)

Potential Reactors Needing Best High Low .Best High Low

-Modifications Modifications Estimate Estimate Estimate Estimate Estimate Estimate

1. Assess plant's 67 150 200 100 10,000 13,000 7,000 capability to cope with a station blackout

2. Develcpprocygyres 67 75 100 50 5,000 7,000 3,000 and training

3. Improve diesel 15 250 2,800 250 11,000 19,000 6,000 generator h reliability (4) '

1

, 4. Increase capability 10 1,000 2,000 500 10,000 20,000 5,000 o to' cope with a h station blackout j (i.e., increase ;

i' capability of station '

batteries, condensate .

storage tank, and/or -

i

!- instrument air)

...., , . . i r. , ,e,

!! 5. Add an AC-independent --

3,000 5,000(6) 2,000 , - - -- --

charging pump (non-seismic) capable 'n . '

, of delivering 50 to 100 . . r ' -

gpmtoreayggrcoolant i. , >

pump seals i'

i Totals 36,000 59.000 21,000 -t

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. , }

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Footnotes for Table 6 (1) Based on 67 reactors. The value, in terms of risk reduction, is also estimated in Section 4.a for a total of 67 reactors. Costs associated with increased reporting requirements for industry and for review by NRC staff are small compared to other costs. These costs are estimated to be about $10,000 and $7,000 per plant, respectively.

(2) Cost estimates are based on informal contacts with utility and architect / engineering firms and should be-viewed as preliminary estimates. Firm cost estimates would require plant-specific designs

- fo.r estimating actual equipment and associated installation costs.

Detailed independent cost estimates presented in Reference 10 indicate icwer costs than shewn above for modification 5 and comparable costs for modifications in item 4 in the table.

(3) Estimates are based on information presented in Reference 11 on-procedures and training for DC power systems. .

5 :: -

(4) The icw estimate per reactor includes a reliability investigation (estimated $100,000) and hardware costs (estimated $150,000). The high estimate per reactor assumes, in addition, plant shutdown for -

five days to requalify a diesel generator at an estimated cost of

$500,000 per day for replacement power. However, not all. plants with low diesel generator reliability would need to shutdown to requalify a~

diesel generator. The basis for the icw, best and high estimated total costs was that one, three or six plants out of an estimated fifteen plants, respectively, that need to improve diesel generator reliability would shutdown for five days to requalify a diesel generater.

1

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_ . . . _ _ _ _ - ~ . _ . _ _ . . _ , . _ . . . __ _ _ _ . _ __ . _ _ ..___

Footnotes for Table 6 (Continued)

(5) It is assumed that reactor coolant pump seal integrity is sufficient to assure core cooling for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or more, and therefore the charging pump would not be necessary. The results of Generic Issue B-23 will provide detailed information on expected pump seal behavior without seal cooling (See Section 4.b.(iii)'for further discussion.) Estimated costs

~-

are provided here for perspective should such a system be considered necessary after B-23 results are available.

(6) A seismically qualifikd and' safety grade AC-ind'ependent charging pump would be much more expensive ($10 - 15 million estimated cost) and would not reduce the risk substantially more than a non-seismic pump.

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c.. . . . . . ... z ..-.

l 1

V = NAFU l op and U = C/m [(e -rt i)/r2] [1 - e -r(t f-t 9)3(7 ,-rm) where

- V,p = value.of avoided onsite property damage N = number of affected facilities = 67

-5

._.aF. = reduction in accident frequency = 3 x 10

"[_ IU' = present value of onsite property damage . ._

C = cleanup and repair costs = $1.2 billion t

f

= years remaining until end of plant life = 25

. tg = years.before reactor.begins operation = 0 r = discount rate = .10 (10%)

m = period of. time over which damage costs are paid out (recovery

)

period in years) = 10 Using the above values, the present value of avoided onsite property

damage is estimated to be $15 million. If avoided costs for replacement power are included (estimated to be $1.2 billion over 10 years in Reference 12), then the estimated present value is $3?i million. The table below summarizes the discounted present value of avoided onsite property damage for 10% and 5% discount rates.

Table 7. Discounted present value of avoided onsite property damage for 67 reactors Discounted Present Value

10% discount rate 5% discount rate 6 6 Cleanup and repair only $15 x 10 $30 x 10

, Cleanup, repair, and 6 6 replacement power $30 x 10 $60 x 10 e

F

l (c) Value-Impacthatio '

Table 8 provides a summary of the total benefits and costs associated with the proposed resolution discussed in Section 3.a of thic Regulatory Analysis. These include: (1) public risk reduction due to avoided !

1 offsite releases associated with reduced accident frequencies; (2) increased occupational dose from implementation, and operation and . -

maintenance activities, as well as reduced occupational exposure from -

cleanup and repair because of lower accident frequency; (3) industry costs for implementation of modifications, operation 'and maintenance, and increased reporting requirements; and'(4) NRC ' costs for rev'ew f of industry reports.

The estimated total cost for industry to comply with, the proposed rule is $40 million. The total public risk reduction for 67 reactors over the remaining life of the plants is about 80,000 person-rem. The overall -

value-impact ratio, not including accident avoidance costs,-is about 2,000 person-rem averted per million dollars. If cost savings to industry from accident avoidance (cleanup and repair of onsite damages and replacement power) were included, the overall value-impact ratio would improve significantly. At a 10', discount rate, the present value of avoided cleanup, repair and replacement power is approximately $30 million. If this benefit were taken into account, the overall value-impact ratio would be about 8,000 person-rem averted per million

dollars.

I (d) Special Considerations The quantitative value-impact analysis discussed above utilized "best l estimates" for benefits (risk reduction) and costs associated with the j proposed resolution of USI A-44. While this is a useful approach to

, evaluate the proposed resolution, other factors can and should play a part in the decision-making process. Although not quantified, other considerations that bear on the overall conclusions and proposed recommendations to resolve USI A-44 are discussed below. Some ,

t _. _ - . - . _ _ _ _ . - . - - 2 _ ____. . _ - . _ _ . _ _ . = ._ ._ ..___ i I

I i

Table 8. Value Impact Summary for Proposed Station Blackout Resolution

. - - . Dose Reduction (person-rem) - Cost ($1,000) -

Best High Low Best High Low Attribute Estimate Estimate Estimate Estimate Estimate Estimate Public Health 80,000 Occupational Ex osure (Accidental)(I 2,000 Occupational Exposure .

(Routine)I2) NA Industry Implementation 36,000 59,000 21,000 Industry Operation (3) 2,000 3,000 1,000 NRC Implementation I4) . . 500 500 500 Total 82,000 38,500 62,500 22,500 Value-ImpactRatio(5): Public dose reduction '

divided by sum of NRC and industry costs (person-rem /$106 ) 2,100 1,300 3,600 1

8ased on an estimated occupational radiation dose of 40,000 person-rom for post-accident cleanup and repair activities (Reference 8). ~

2 No significant increase in occupational exposure is expected from operation and maintenance or implementing the recommendations proposed in' this resolution. Equipment additions and modifications contemplated do not require significant work in and around the reactor coolant system and therefore would not be expected to result in significant radiation exposure. NA = not affected.

3 Assumes 5% of installation costs for operation and maintenance.

4 Based on an estimated 120 person-hours per reactor for NRC review.

5 This does not take into account the additional benefit associated with avoided plant damage costs or replacement power costs resulting from reduced frequency of core melt. The cost for plant cleanup following a core melt accident is estimated to be $1.2 billion, and replacement power is estimated to cost about 5500,000 per day (Reference 8). The estimated discounted present value of these avoided onsite costs is given in Table 7.

emuD y . - - -

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of these factors mai indicate that the proposed resolution is conservative (e.g., possible reduction in source term), while others may show that the resolution is non-conservative (e.g., median core melt

~

frequencies were used instead of means). Overall, these considerations support the conclusion that additional defense-in-depth provided by the

, ability of a ' plant to cope with a station blackout for a speciff'ed duration is strongly recommended.

~

Relative Importance of Potential Station Blackout Events Probabilisticriskassessmentstudies(PRA)performedforthisUSI,as well as some plant-specific PRAs, have shown that station blackout can be a significant contributor to core melt frequency, and with the

~

consideration of containment failure, station blackout events can

! represent an important contributor to reactor risk. In general, active containment systems required for heat removal, pressure suppression, and radioactivity removal from the containment atmosphere following an accident are unavailable during a station blackout. Therefore, the offsite risk is higher from a core melt resulting from station blackout than it is from many other accident scenarios.

Source Term Re-Evaluation The consequence estimates for different containment types used in this value-impact analysis are consistent with the guidance given in NRC's value-impact handbook (Reference 12). However, substantial uncertainties exist regarding fission product transport in containment during a core melt accident. For example, the time to containment failure following the onset of core damage and the containment failure l mode represent important factors which influence fission product release, and ultimately public risk. Ongoing research on severe l accidents and source term re-evaluation may significantly alter estimated fission product releases in the future and thereby reduce the ~

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estimated risk to the public. In any event, the ability to cope with a

, station blackout for some period of time would make station blackout a 4 small contributor to core melt frequency and would significantly reduce the risk associated with such events.

J'

Mean or Median Core Damage Frequencies The estimates of core damage frequencies for this value-impact analysis

! were calculated using " point estimates" of accident frequencies (Reference 5). This is intended to be a realistic analysis (e.g.,

~

tending to be neither" conservative nor non-conservative) in tha't there j is approximately an equal probability of plant core damage frequencies l being higher or lower than this value. Mean values of core damage frequencies, which are higher than the point estimates, could have been used. However, the calculation of mean values depends on the assumed j distribution (e.g., log normal or other) for probabilities' of event frequencies which in itself leads to some uncertainty (Reference 4).

Future Trends in Loss of Offsite Power Frequency I

The estimated frequency of core damage from station blackouts events is

! directly proportional to the frequency of the initiating event.

Estimates of station blackout frequencies for this USI were based on j actual operating experience. This is assumed to be a realistic'-

~

indicator of future performance. An. argument can be made that the future perfonnance will be better than the past. For example,'when problems with the offs:te power grid arise, they are fixed, and therefore, grid reliability should improve. On the other hand, grid power failures may become more frequent because fewer plants are being

built, and more power is being transmitted between regions, thus l ' placing greater stress on trans;nission lines.

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Comon Cause Failures One factor that affects AC power system reliability is the vulnerability to common cause failures associated with design, operational, and environmental factors. Comon cause failure vulnerabilities are plant-specific; therefore, generic requi.rements to improve the reliability of AC power systems by eliminating specific single-point vulnerabilities is not practical. Existing industry and NRC standards and regulatory guides include specific design criteria and guidance on the independence of offsite power circuits and the -

independence of, and ' limiting interactions between, diesel generator units at a nuclear station.. In developing the proposed resolution of USI A-44, the NRC staff assumed that, by adhering to such standards, licensees have minimized, to the extent practical, s. ingle-point vulnerabilities in design and operation that could result in a loss of all offsite power or all onsite emergency AC power. Results of sensitivity studies presented in Reference 5 indicate that if potential common cause failures of redundant emergency diesel generators exist (e.g., in service water or DC power support systems), then estimated core damage frequencies can increase significantly.

4 Alternative (ii) -

Treat All Plants Uniformly The alternative of treating plants uniformly by requiring all plants to be able to cope with the same station blackout duration has been considered. This simplified approach has the advantage of being potentially easier to implement, but it also has two major drawbacks.

First, operating nuclear power plants have significant differences in plant- and site-specific factors that contribute to risk from station blackout. This alternative would not take these known factors into account. For example, plants that have a more redundant emergency AC power system than other plants would not be given credit for such

features. Second, requiring all plants to be able to cope with the same blackout-duration would result in one of two undesirable alternatives:

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(1) If a uniform duration of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or less were recommended, station blackout could still be a significant contributor to total core melt frequency for some plants; and therefore,~ the objective of the proposed requirements would not be met; (2)'If a uniform 8-hour requirement were imposed, it would necessitate expenditures at some plants that would not be considered cost-effective in reducing the risk from station blackout ev'ents. Therefore, this alternative is not recommended.

Alternative-(iii)

Another possible itlternative to the recommended action is'to require '

plants to install either an additional emergency diesel generator or another AC-independent decay heat removal system. This alternative is not being recommended for several reasons. First, the cost for either of these additions (from $10 to $30 million per plant) would be much higher than the estimated cost for the proposed resolution. The

. reconnended approach is more cost effective and meets the objective stated in Section 2. Second, the adequacy of present requirements for decay heat removal systems is being studied under USI A-45, and any major hardware changes or additions to these systems should await the technical resolution of A-45. Third, experience. indicates that .there are practical limits to diesel generator reliability, including common cause:

failures of redundant divisions, and the reconnended resolution provides greater diversity and additional defense-in-depth. ,

Alternative (iv) -

Take no action.

This alternative would be to take no actions beyond those resulting from the proposed resolution of Generic Issue B-56 (discussed in Section 4.b.1 of this Regulatory Analysis). Operating ' experience with diesel generator failures and losses of offsite power has raised a significant concern regarding the potential risk from a station blackout event.

The use of this' da.ta base with relatively straightforward application of probabilistic risk assessment techniques indicates that station blackout eW

events could be a significant contributor to risk for many plants. The additional actions recommended for USI A-44 would significantly reduce the estimated frequency of core melt associated with severe accidents from station blackout. Since the value-impact analysis has shown that it would be beneficial to implement the proposed recommendations, the no-action alternative is not recommended. ,

b. Impacts on Other Requirements There are several ongoing NRC generic programs and requirements that are related to t'he USI A-44 activities and proposed regulatory positfons.

These items are discussed on the following pages.

(1) Proposed Actions for Enhancing Reliability of D.iesel Generators at Operating Plants, Generic Issue B-56 The Nuclear Regulat'ory Comission has requested information from

~

licensees regarding proposed actions to improve and maintain diesel generator reliability (Generic Letter 84-15, July 2,1984). The letter requested specific information on the following three areas:

(a) Reduction of cold fast-start surveillance tests for diesel generators.

(b) Ofesel generator reliability data.

(c) Description of licensee's diesel generator reliability program, if any, and comments on the staff's example performance technical specifications for diesel generator reliability.

After. staff review of responses to the request for reliability data, those plants whose diesel generator reliability is below a i

I -

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- ...: . . .: . :.: - u = .:.- -

i i

minimum specified value will be considered on'a case-by-case basis

to improve diesel generator performance.

The example performance specifications, item (c) above, would establish a set of action statements based on the reliability actually exhibited by diesel generators with 0.05 failures per demand as the maximum desired failure rate and 0.10 failures per i demand as the maximum acceptable failure rate. Licensees with a I

diesel generator failure rate above the desired level would have to !

increase testing frequency from once every 31 days to once every 7

~

days'. This test frequency'would b'e maintained until seven consecutive failure-free demands have been performed and the number of failures in the last 20 demands has been reduced to one or less.

Licensees with diesel generator failure rate above the maximum f acceptable level would be. required to disqualify their diesel generators. The licensee would initiate a requ'alification program to demonstrate that the causes of the failure have been corrected.

j The diesel generator would be requalified only if the failure rate has been reduced to 0.05 failure per demand or less. The testing l frequency and the allowable out-of-service period would also be i .

relaxed somewhat from that currently recommended by Regulatory i Guide 1.108.

The example perfonnance specifications are compatible with the proposed requirements to resolve USI A-44. In fact, the proposed resolution of A-44 assumes that responses to the generic letter, including licensee comments on the example performance technical specifications, will establish a basis for maintaining and determining the reliability of emergency diesel generators. The i

example program is based on a desired reliability level of 0.05 s

failures per demand for all plants, whereas, the resolution of USI A-44 proposes maintaining failure rates at or below 0.05 or 0.025 i

.. _ - - - - - . - - . . ~ . , , . .-. ., ,,

failures per demand depending on the plants' characteristics as discussed in Section 3 of this Regulatory Analysis.

(ii) USI A-45, " Shutdown Decay Heat Removal Requirements" The overall objective of USI A-45 is to evaluate the adequacy of current licensing design requirements in order to ensure that nuclear power plants do not pose an unacceptable risk due to failure to remove shutdown decay heat. The study includes an assessment of alternative means of shutdown decay heat removal and '

of diverse " dedicated" systems for this purpose. Results will include proposed recommendations regarding the desirability of, and possible design requirements for, improvements in existing ystems or an alternative decay heat removal method. .

The U.SI A-44 concern for maintaining adequate core cooling under station blackout conditions can be considered a sub-set of the overall A-45 issue. However, there are sigr.ificant differences in scope between these two issues. USI A-44 deals with the probability of loss of AC power, the capability to remove decay heat using systems that do not require AC power, and the ability to restore AC power in a timely manner. USI A-45 deals with the overall reliability of the decay heat removal function in terms of response to transients, small break loss of coolant accidents and -

special emergencies such as fires, floods, seismic events and sabotage.

Although the recommendations that might result from the . . _ acion of USI A-45 are not yet known, some could affect the station blackout capability, while others would not. Recommendations that involve new or improved decay heat removal systems that are AC power dependent but that do not include a separate independent AC 5

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l power supply would have no effect on USI A-44. Recommendations that involve an additional AC-independent decay heat remeval system would have a very modest effect on USI A-44. Recommendations that involve a new decay heat removal system which would include an independent AC power supply would have a significant effect on USI A-44. Such a new additional system would receive the appropriate credit with'in the present structure of the proposed USI A-44

, resolution by changing the Emergency AC Power Configuration Group.

The proposed resolution of USI A-44 would necessitate expenditures typically around $C00,000 per plant with a range estimated to be from a few hundred thousand dollars to a naximum of around $4 million. A resolution for USI- A-45 involving a new additional comprehensive system, such as an additional shutdown cooling system with its own dedicated diesel generator, would be much more expensive involving an expenditure on the order of tens' of millions of dollars.

The resolution of these two issues is coordinated along two main lines. First, technical information resulting from both studies is shared among the major participants including NRC staff ar.d contractors. In this way the resolution of USI A-45 will take into account any modifications resulting from the resolution of USI A-44 -

that are applicable to the decay heat removal functioncrSecond, -

the schedules are coordinated such that.by the time a final rule is

~

l published on USI A-44, and well before plant modifications, if any, would be implemented, the proposed technical resolution of USI A-45 will be published for public comment.

The draft contractor technical report for USI A-45 is scheduled to be completed in early 1985, and the draft regulatory position of the proposed resolution of USI A-45 is scheduled for the Spring of O

me

. . . . -1

- 34 -

l 1985. If the rule on USI A-44 were issued, it would not be final

.until the Spring of 1986. For plants needing hardware modifications to comply with the proposed USI A-44 resolution, this 4

schedule will permit a re-evaluation before any actual modifcations are made so that any contemplated design changes following from the resolution of USI A-45 can be considered at the same time.

(iii) Generic Issue B-23, " Reactor Coolant Pump (RCP) Seal Failures" .

The Task Action Plan for B-23 identifies several tasks to resolve

~

this issue including a . review of seal failure opera' tin'g experience, an assessment of the effects of loss of seal' cooling on reactor coolant pump (RCP) seal' behavior, and an evaluation of other causes of RCP seal failure such as mechanical and maintenance-induced failures.- Work by NRC and industry is proceeding to analyze seal

-performance with loss of seal cooling. Since reactor coo.lant pump seal integrity is necessary for maintaining primary system inventory under station blackout conditions, the results of this analysis will provide information to determine seal behavior and, likewise, a plant's ability to cope with a station blackout for a i

specified time.

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wp

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(iv) Gbneric Issue A-30, i' Adequacy of SafetIy-Related DC Power Supply" The staff's proposed resolution of this issue specifies guidance for enhancing the reliability of DC power supply systems. This "

guidance includ'es items such restricting interconnections between i redundant DC divisions, monitoring the readiness of the DC power

system, and specifying administrative procedures and . technical specifications. for surveillance testing and maintenance activities. _

The analysis performed for USI A-44 (Ref. 5) assumed that a high Tevel of DC power 'sys-tem reliability wou'1d be' maintained'so that (1) DC power system failures would not be a significant contributor to losses of all AC power, and (2) should a station- blackout occur,-the probability of immediate DC power system failurg would be low.

Whereas the proposed resolution of Generic Issue A-30 focuses on enchancing battery reliability, the proposed resolution of USI A-44 is aimed at assuring adequate station battery capacity in the

event of a station blackout of a. specified duration. Therefore, the proposed resolutions of these two issues are consistant and

' compatible.

(v) Regulatory Guide 1.108, " Periodic Testing of Diesel Generator l Units Used as Onsite Electric Power Systems at Nuclear Power Plants" This regulatory guide describes the currently acceptable method for complying with-the Commission's regulations with regard to periodic testing of diesel generators to ensure that they will meet their availability requirements. This guide may need to be modified to be consistent with the proposed actions described in item 1 above (Generic Issue B-56). If necessary, Regulatory Guide 1.108 will be revised:after comments have been received from licensees on the B-56 example diesel generator reliability program.

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(vi) Fire Protection Program for Nuclear Power Facilities

,10 CFR 50.48 states that each operating nuclear power plant shall have a fire protection plan that satisfies Criterion 3 of the General Design Criteria. The fire protection features required to satisfy Criterion 3 are specified in Appendix R to 10 CFR 50 and BTP CMEB 9.5.1, and include certain provisions regarding

alternative and dedicated shutdown capability'. To meet these r -

provisions, some facilities have added, or plan to add,' improved capability to restore power from offsite sources or onsite diesels for the shutdown systent. A few plants 'h' ave installed a safe shutdown facility for fire protection that includes a charging pump

powered by its own independent AC power source. In the event of a station blackout, this system can provide makeup cap, ability to the

primary coolant system as well as reactor coolant pump seal 1

cooling. This could be a'significant benefit in terms of enhancing the ability of a plant to cope with a station blackout.

l Since the plant modifications required for fire protection have i

already been specified, it would not be feasible to consider these modifications together with the proposed requirements of USI A-44.

I However, credit is given for improvements made for the fire protection program in the USI A-44 resolution. --

_ (vii) Proposed Auxiliary System Reliability and Power Diversity Upgrade in Operating Plants l

l The Division of Systems Integration is considering a proposal ht'at

! the reliability' and diversity requirements for auxiliary feedwater systems specified in Standard Review Plan (SRP) Section 10.4.9, -

" Auxiliary Feedwater System," be applied to all PWRs. It is planned to submit this proposal to the CRGR_for review. This proposed requirement, if adopted, would require an upgrading of the i i

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auxiliary feedwater pump train in several PWRs. This upgrading would most likely result in the addition of an auxiliary feedwater pump. Since all PWRs new have an auxiliary feedwater system that is AC-power independent, the criteria specified in SRP 10.4.9 could be met by adding a motor-driven pump. Consequently, the auxiliary feedwater system upgrade could have no effect on the station blackout issue. T'he installation of a system that is independent of AC power would be beneficial in handling station blackout accident sequences. Since this issue is so closely related to USI

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A-45, " Shutdown Decay Heat Removal Requirements," its resolution

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may await the reconsnendations' of USI 'A-45.

(viii) Multiplant Action Items B-23 and B-48, " Degraded Grid Voltage" and

" Adequacy of Station Electric Distribution Voltage" These two multiplant action items. have been under consideration by both the staff and licensees for several years. They relate to:

(1) sustained degraded voltage conditions at the offsite power sources, (2) interaction between the offsite and onsite emergency power systems, and (3) the acceptability of the voltage conditions on the station electric distribution systems with regard to potential overloading and starting transient problems. Licensee responses to these concerns have consisted of verifying the adequacy of existing power systems or of upgrading the power systems. These modifications are designed to assure that the power systems can perfonn their intended function and consequently would enhance their dependability. If additional power sources have been added to address these concerns, the plant would be placed in an improved category and may be required to withstand a blackout of lesser time duration. In the resolution of USI A-44, we are not recommending that work that has been done on these two action items be repeated.

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c. Constraints ,

The staff has reviewed current Comission regulations to determine if ,

they provide a basis for implementation of the proposed USI A-44 requirements. This review included: (1)theAtomicSafetyand Licensing Appeal Board Hearing, ALAB-603, on station blackout for St.

Lucie Unit 2, (2) the Commission review of that hearing, and.(3) General Design Criterion 17. " Electric Power Systems." A sumary of this review ,

is presented belo'v.

'(i) 'ALAB-603 took the position tha't station blackout'should be considered a design basis event for St. Lucie Unit 2 because of the unacceptably high frequency of such an event (10-4 to 10-5 per year at that site). As a result,.the Appeal Board required St. Lucie 2 to be capable of withstanding a total loss of AC p6wer and to implement training and procedures to recover from station blackout. The Appeal Board went as far as to say:

"Our findings that station blackout should be considered as a design basis event for St. Lucie Unit 2 manifestly could be applied equally to Unit 1, already in operation at that site.

By a parity of reasoning, this result may well also obtain at

. other nuclear plants on applicant's system, if not at most -

power reactors. Our jurisdiction, however, is limited to the matter before us, licensing construction of St. Lucie 2.

Beyond that, we-can only alert the Comission to our ;

concerns."

(ii) The Comission upheld the Board's action on St. Lucie 2. ;

However, the Comission determined that ALAB-603 did not l establish station blackout generically as a " design basis l

event."

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(iii) The last paragraph of GDC 17 states that: " Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies."

GDC 17's intent is to require reliable offsite and onsite AC power systems. The ability to cope with the coincident loss of both of these systems is notddressed a explicitly.

As a result of this review, the staff has concluded that there is a basis in the regulation for the proposed recommendations to improv,e the reliability of the offsite and onsite AC power systems. However, since the coincident loss of both systems is not addressed explicity, we are proposing a rule to require plants to be able to withstand a total loss of AC power for a specified duration in order to provide further assurance that station blackout will not adversely affect the public health and safety.

No other constraints have been identified that affect the proposed resolution of USI A-44.

5. Decision Rationale -

The evaluation to resolve USI A-44 included deterministic and probabilistic analyses. Calculations to determine the timing and consequences of various accident sequences were performed, and the dominant factors affecting station blackoutlikelihoodwereidentified(References 2-5). Using this information, simplified probabilistic accident sequence correlations were calculated to estimate the likelihood of core melt accidents resulting from station blackout for different plant design, operational, and location factors. These quantitative estimates were used to give insights on the

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relative importance of various factors, and those insights, along with engineering judgement, were used to develop the proposed resolution. By analyzing the effect of variations in design, operations and plant location on risk from stat' ion blackout accidents, an attempt was made to approach a reasonably consistent level of risk in the recommendations developed.

A survey of probabilistic risk assessment studies showed that total core damage frequency from all dominant accident sequences ranged from 2 x 10-5 to 1 x 10-3 per reactor-year with a typical frequency of about 6 to 8 x 10-5 p,7 .

reactor-year (Reference 4). For plants in operation or under construction, a .

value-impact analysis was performed to determine that the proposed recommendations to resolve USI A-44 were cost-effective. The proposed resolution would result in station blackout being a relatively small contributor to core damage frequency. A more detailed discussion of the analysis of station blackout accident likelihood performed for this

. Regulatory Analysis is provided in Reference 5.

Additional rationale for recommending implementation of the proposed rule and the draft regulatory guide (Reference 6) over other alternatives is included in the value-impact analysis discussed in Section 4.a of this Regulatory Analysis. This proposed action represents the staff's position based on a comprehensive analysis of the station blackout issue. This position includes ~

all the requirements and guidance to resolve the station blackout issue. - - -

The studies and data on which these recommendations are based are documented in References 2 - 5. Summaries of the three contractor NUREG reports (Ref. 2

- 4) and the staff NUREG report (Reference 5) are presented below.

(a) Summary of NUREG/CR-3226, " Station Blackout Accident Analyses" This report analyzes accident sequences following a postulated total loss of AC power to (a) determine the core damage frequencies from station blackout, (b) provide insights through sensitivity studies of important factors to consider for icwering the core melt frequency, and (c) provide perspectives i

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on the risks from such an event. Probabillstic safety analyses were done on four generic " base" plant configurations. Fault trees of different systems and event trees of possible station blackout accident sequences were constructed for these plants. These event trees modeled three time periods including an initial time period for sequences resulting from unavailabilities on demand and longer time intervals in which other failures can occur such as depletion of DC power, degradation of RCS pump seals, or depletion of condensate storage tank supply. Data from the offsite and onsite power studies (References 2 and 3) as well as from LERs and PRAs were

~

used to quantify the accident sequences. Lastly, containment failure modes and timing were reviewed in ordir to calculate the risk to the public ffom station blackout.

For the " base" cases, the total core melt frequencies fro,m station blackout resulting from the dominant accident sequences were estimated to be in the range of 10-5 per reactor year. Plants with features different from the base case designs have different core melt frequencies, so sensitivity analyses were conducted. For example, the reliability and recovery of AC power from both the offsite and emergency onsite power systems have a direct impact on core melt frequencies. Depending on the expected frequency of station blackout at a plant and other factors, the frequency of core melt associated with loss of all AC power ranged from about'2 x 10-6 to greater than 10-4 per reactor-year. - -

In summary, results of the accident sequence analyses indicate that the following plant factors are important when considering station blackout:

The effectiveness of actions to restore offsite power once it is lost.

The degree of redundancy and reliability of the emergency onsite AC power system.

The reliability of decay heat removal systems following loss of AC power.

DC power reliability and battery capacity including the availability of instrumentation and control for decay heat removal without AC power.

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Common service wa'ter dependencies between the emergency AC power source and the decay heat removal systems.

The magnitude of reactor coolant pump seal leakage and the likelihood of a stuck open relief valve during a station blackout.

Containment size and design pressure.

Operator training and available procedures.

(b) Sunnary o'f NUREG/CR-2989,'" Reliability of Emergency AC Power Systems at -

Nuclear Power Plants" The purpose of this s'tudy 'was to estimate the reliabi11 ties of representative onsite AC power systems and to estimate the costs of fixes to improve the re11 abilities of these systems. For this analysis, an initial design review of onsite AC power systems was done using FSARs, plant schematics, and plant-specific procedures. The study included examining the following areas:

switchyards, distribution systems, DC power systems, diesel generators, support systems and procedures. Historical data on diesel generator.

operating experience for the five-year period from 1976 through 1980 were collected from licensee event reports and responses to questionnaires sent to -

nuclear plant licensees.

- Eighteen different configurations were identified, and . representative plants p were selected for a more detailed reliability analysis. This analysis

-involved constructing fault tree models for the onsite power systems and -

i quantifying these fault trees .with the data gathered on operating experience.

The onsite system tr4ependability, the probability that it will fail to start or fail to continue to run for the duration of an offsite power outage, was calculated for AC power outages up to 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months after a loss of offsite power.

l Results of a sensitivity study were used to identify potentially important contributors to the unreliability, and costs of improvements were estimated.

Results showed that important contributors to onsite power undependability l

were independent diesel generator failure, common cause failure due to hardware failure or human error, unavailability because of scheduled i

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. - - . . . - . . . . - . - . . - . . . . . - . . . . . . . . - . - - - . . . _ . . . - .. - - - -.--...w maintenance, and cooling subsystem undependability. ' Reliabilities of onsite AC power systems vary from plant to plant. Depending on diesel generator configuration, the system unavailability ranged from 1.4 x 10-4 to 4.8 x 10-2 per demand. Significant variability exists so that any reliability improvements and the associated costs must be evaluated on a plant-specific basis.

(c) Sumary of NUREG/CR-3~992, " Collection and Evaluation of Complete and ,

Partial Losses of Offsite Power at Nuclear Power Plants"

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This report describis'and categorizes events involving complete or

significant partial losses of offsite power that have occurred at nuclear power plants through 1983. The purposes of this study were to provide an accurate data base to estimate fr,equencies and durations of losses of offsite power and to understand how offsite power design features may affect these losses as well as the ability to restore offsite power.

Based on industry-wide data for the years 1959 through 1983, the frequency of loss of offsite power is about once every 10 site-years. A total of 46 complete loss of offsite power events were documented ranging in duration from a few minutes up to a maximum of almost nine hours. In approximately half of these events, offsite power was restored in one-half hour or less.

Infonnation for this study was collected from licensee event reports (LERs),

responses to an NRC questionnaire, and various reports prepared by the utilities. Most of the event descriptions contained in the LERs and other documentation within the Nuclear Regulatory Commission ~ files did not contain sufficiently detailed information for the purposes discussed above. For example, in one case a licensee reported offsite power restoration time to be 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , but actually one offsite power source was restored in 8 minutes, and all offsite power was restored in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Since restoration of one source of offsite power terminates a loss of offsite power, the documented description was not accurate enough. In some other cases, offsite power was ;

available to be reconnected, but the plant operators did not reconnrct it for I some time after it was available. The time power was reconnected was usually I

, reported in the LERs, however the data that were actually needed were the ,

i

times that power was available for reconnection. Because of the need for more accurate data, additional information was obtained by contacting utility engineers to obtain better descriptions of the causes, sequences of events, and the times and methods of restoring offsite power.

Once these data were collected, the offsite power failures were identified as

' plant-centered or grid failures. In addition, the causes of the failures were attributed to weather, human error, design error,-or hardware failure.

The plant-centered failures were usually of shorter duration than the grid failures caused by severe weather. For this reason, the weather-related events weie reviewed in detail.

Offsite power design features were tabu' lated for most of the operating nuclear power plants to determine which ones significantly affect offsite power system reliability. The frequency and duration of losses of offsite power caused by severe weather are affected by the number of transmission lines and rights-of-way and the availability of alternate power sources, such as hydro, gas turbines, or fossil units near the nuclear plant. Design features that may be important for plant-centered losses of offsite power are I

the number of offsite power sources, the electrical independence of those sources, and the relay scheme for transferring power between offsite sources.

(d) Sumary of NUREG-1032, " Evaluation of Station Blackout Accidents at ;

Nuclear Power Plants, Technical Findings Related to Unresolved Safety Issue A-44" This report sumarizes the results of technical studies performed in support

! of USI A-44 and identifies the dominant factors affecting the likelihood of station blackout accidents at nuclear power plants. These results are based j on operating experience data; analysis of several plant-specific probabilistic safety studies; and reliability, accident sequence, and consequence analyses performed in support of this unresolved safety issue.

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In suninary the results show the following important characteristics of station blackout accidents.

(1) The' likelihood of station blackout varies between plants with an estimated frequency ranging from approximately 10-5 to 10-3 per reactor-year. A " typical" estimated frequency is on the order of 10-4

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_ per reactor year.

(2) The capability of restoring offsite power in a timely manner can have a significant effect on accident consequences. ,

(3) The onsite AC power system redundancy and individual power supply reliability have the largest influence on station blackout accident frequency. .

- (4) The capability of the decay heat removal system to cope with long duration blackouts can be a dominant factor influencing the likelihood of core damage or core melt, i

(5) The estimated frequency of station blackout events resulting in core i damage or core melt can range from approximately 10-6 to greater than 10~4 per reactor-year. A " typical" core damage frequency estimate is on '

the order of 10-5 per reactor-year. -

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(6) The best information available indicates that containment failure by '

l overpressure may follow a station-blackout-induced core melt with

! smaller, low design pressure containments most susceptible to early j failure. Some large, high design pressure containments may not fail by i overpressure, or the ' failure time could be on the order of a day or more.

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It was found that losses of offsite power could be characterized as those resulting from plant-centered faults, utility grid blackout, and severe weather-induced failures of offsite power sources. The industry average frequency 'of total losses of offsite power was determined to be about one in 10 site-years. The median restoration time was about one-half hour, and 90%

of the losses were restored in three hours or less. The factors which were identified as affecting the frequency and duration of offsite power losses -

are:

(1) Design of preferred power distribution system, particularly the number and independence' of offsite power circuits from the point where they enter the site up to the safety buses; (2) Operations which can compromise redundancy or independence of multiple offsite power sources, including human error; I

(3) Grid stability and security, and the ability to restore power to a nuclear plant site with a grid blackout; and (4) The hazard from, and susceptibility to, severe weather conditions which can cause loss of offsite power for extended periods.

A design and operating experience review, combined with a reliability .-

analysis of the oasite, emergency AC power system, has .shown that there are a + -

variety of potentially important failure causes. The typical unavailability of a two-division emergency AC power system is about 10-3 per demand, and the typical individual emergency diesel generator failure rate is about 2 x 10-2 per demand. The factors which were identified as affecting the emergency AC power system reliability during a loss of offsite power are:

(1) Power supply configuration reiundancy; l

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(2) Reliability of each power supply; (3) Dependence of the emergency AC power system on support of auxiliary cooling systems, control systems; and (4) Vulnerability to common cause failures associated wi,th design, operational and environmental factors.

The likelihood of a station blackout progressing to core damage or core melt is dependent on the reliability and capability of decay heat removal systems which are not dependent on AC power. If sufficient capability exists, additional time will be available to permit an adequate opportunity to restore AC power to the many systems normally used to cool the core and remove decay heat. The most important factors involving decay heat removal during a station blackout are:

(1) The starting reliability of systems required to remove decay heat and ,

maintain reactor coolant inventory; (2) The capacity and functionability of decay heat removal systems and auxiliary or support systems which must remain functional during a station blackout (e.g., DC power, condensate storage);

(3) For PWRs, and BWRs without reactor coolant makeup capability-during a +~ ~

station blackout, the magnitude of reactor coolant pump seal leakage; and (4) For BWRs which remove decay heat to the suppression pool, the ability to maintain suppression pool integrity and operate heat removal systems at high pool temperatures during recirculation.

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. l It was' determined by reviewing design, operational, and location ' factors, that the expected core damage frequency from station blackout could be maintained around 10-5 per reactor-year or lower for all plants. The ability to cope with station blackout durations of 4 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and emergency diesel generator reliabilities of 0.95 per demand or better would be necessary to reach this core melt frequency level.

6. Implementation _ _ _ . . _ _

(a) Schedule for Implementing the Proposed Rule The steps and schedule listed below are suggested for implementing the proposed rule. Within 9 months after final promulgation of the rule, licensees will submit to NRC the duration for which the plant should be able to cope with a station blackout, a justification for the duration, and,a description of the procedures to cope with a station blackout for that duration. The NRC staff will review the licensees' submittals, and within 6 months after that review licensees will submit a schedule for implementing equipment modifications, if necessary, to comply with the proposed rule. The factors that need to be considered to determine the minimum acceptable station blackout duration, as specified in the

proposed revision to Appendix A, GDC 17, are relatively simple. Thus, this duration can be determined in approximately one or two months. -

Licensees will be required to perform plant-specific analyses to determine how long the plant, as designed, can cope with a station blackout, and to determine what modifications, if any, are needed to meet the acceptable duration. These analyses could require 6 to 9 months to perform. Thus, it does not seem unreasonable to request that the information be submitted to the NRC within nine months after the date of the proposed rule. A final schedule for implementation of l

I

design and/or procedural modifications will be mutually agreed upon by the licensee and NRC staff. The implementation of procedural changes to cope with a station blackout and diesel generator reliability improve-ments, if necessary, should be accomplished early in the schedule.

Hardware backfits, if necessary, should be implemented as soon as practical based on scheduled plant shutdown, but no later than 2 years after NRC reviews the licensees' station blackout duration submittal.

Months After Submitting Activity Rule to Commission Subm1t proposed rule to Commission 0 Issue proposed rule and regulatory guide for public comment 3 Submit final rule to Commission 13 Issue final rule 16 Licensees submit station blackout duration to NRC, including description of procedures 25 NRC completes review of submittal 36 Licensees complete hardware modifications Schedule to be agreed upon with NRC but within 2 years of NRC review of submittal unless justification is subm.itted by the licensee for a -later date and the staff-agrees.

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Other schedules we're considered; however, we believe the 'above proposed' schedule is achievable without unnecessary financial burden on the licensee for plant shutdown, and allows reasonable time for the implementation of necessary hardware items to achieve a~ reduction in the risk of severe accidents associated with station blackout, yet achieves significant benefits early on by requiring an assessment of a plant's ,

station blackout capability and procedures and training'to cope with such an event. Shorter or less flexible schedules would be - - -

unnecessarily burdensome; longer' schedules would delay necessary plant improvements.

(b) Relationship to Other Existing or Proposed Requirements There are several NRC programs that are related to A,-44; these are discussed in Section 4.b. These programs are compatible with the proposed resolution of A-44.

References

1. WASH-1400, " Reactor Safety Study," October 1975.

2. NUREG/CR-2989, " Reliability of Emergency AC Power Systems at Nuclear Power Plants," July 1983. -

3. NUREG/CR-3992, " Collection and Evaluation of Complete and Partial Losses of Offsite Power at Nuclear Power Plants," to be published.

4. NUREG/CR-3226, " Station Blackout Accident Analyses (Part of NRC Task Action Plan A-44)," May 1983.

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5. NUREG-1032, For Comment, " Evaluation of Station Blackout Accidents at '

Nuclear Power Plants, Technical Findings Related to Unresolved Safety Issue A-44," to be published.

6. Draft regulatory guide, " Station Blackout."

7. - NUREG/CR-2723, " Estimate's of the Financial Consequences of Nuclear Power Reactor Accidents," September 1982.

8. NRR Office _ Letter No.16, Revision 2, " Regulatory Analysis Guidelines,"

October 3, 1984. '

9. Sandia National Laboratory Letter Report to NRC, "Value-Impact Calculation for Station Blackout Task Action Plan A ,44," March 1983.

10. NUREG/CR-3840, " Cost Analysis for Potential Modifications to Enhance the Ability of a Nuclear Power Plant to Endure Station Blackout," July 1984.

11. EG&G Report RE&ET-6151, " Cost Analysis for Enhancement of DC Systems Reliability and Adequacy of Safety-Related DC Power Systems," January 1983.

12. NUREG/CR-3568, "A Handbook for Value-Impact Assessment," December 1983. -

and

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)

e e ENCLOSURE 3 i

REGULATORY GUIDE

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ENCLOSURE 3 DRAFT REGULATORY GUIDE -

STATION BLACK 0UT (TASK RRB- )

A. INTRODUCTION Criterion 17. " Electric Power Systems,." of Appendix A " General Design Criteria for Nuclear Power Plants," to.10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities," includes a requirement that an onsite-electric power system and an offsite electric power system be provided to permit functioning of structures, systems, and components important to safety.

Proposed paragraph (a), " Requirements," of proposed Section 50.63, " Loss of All Alternating Current Power," of 10CFR Part 50, " Domestic Licensing of Production and Utilization Facilities," would require that each light-water-cooled nuclear power plant be able to withstand and recover from a station blackout (i.e., loss of the offsite electric power system concurrent with reactor trip and unavailability of the onsite emergency AC electric power system) for a specified duration. Proposed paragraph (e) of General Design Criterion (GDC) 17 would require that, for the station blackout duration, the plant be capable of maintaining core cooling and containment integrity.

Paragraph (e) would also identify the factors that must be considered in specifying the station blackout duration. The proposed Section 50.63 and the proposed' paragraph (e) of GDC 17 were published in the Federal Register on .

Criterion 18. " Inspection and Testing of Electric Power Systems," of Appendix A to 10 CFR Part 50 includes a requirement for appropriate periodic testing and inspection of electric power systems important to safety.

This guide describes a method acceptable to the NRC staff for complying with j the Commission's proposed regulation requiring nuclear power plants.to be l, l capable of coping with a station blackout for a specified duration. This j l guide applies to all commercial light-water-cooled nuclear power plants.

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i B. DISCUSSION The term " station blackout" refers to the ccmplete loss of alternating current electric power to the essential and nonessential switchgear buses in a nuclear power plant. Station blackout therefore involves the loss of offsite power concurrent with turbine trip and failure of the onsite emergency AC power system. Because many safety systems required for reactor : --..

e core decay heat removal and containment heat removal are dependent on AC- -

power, the consequences of station blackout could be severe. ~In the event of a station blackout, the capability to cool the reactor core would be cependentontheavailabiiityofsystemsthatdonotrequire^Cpowerandon the ability to res' tore AC power in a timely manner.

The concern about station blackout arose because of the accumulated experience regarding the reliability of AC power supplies. Many operating plants have experienced a total loss of offsite electric power, and more occurrences are expected in the future. In almost every one of these loss-of-offsite-power events, the onsite emergency AC power supplies have been available imediately to supply the power needed by vital safety equipment. However, in some instances, one of the redundant emergency AC power supplies has been unavailable. In a few cases there has been a complete loss of AC power, but during these events. AC power was restored in a short time without any serious consequences. In addition, there have been numerous-instances when emergency diesel generators have failed to start and -

run in response to tests conducted at operating plants.

The results of the Reactor Safety Study (Ref. 1) showed that, for one of the two plants evaluated, a station blackout event could be an important contributor to the total risk from nuclear power plant accidents. Although this total risk was found to be small, the relative importance of station

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blackout events was established. This finding and the accumulated diesel generator failure experience increased the concern about station blackout.

In a Comission proceedin~g addressing station blackout, it was determined that the issue should be analyzed to identify preventive or mitigative measures that can or should be taken. (See Florida Power & Light Company 1 (St. Lucie Nuclear Power Plant, Unit No. 2) ALAB-603,12 NRC 30 (1980); . .

modified CLI-81-12, 13 NRC 838 (1981)). -

The issue of station blackout involves the likelihood and duration of the loss of offsite power, the redu'ndancy and reliability of onsite emergency AC power systems, and the potential for severe accident sequences after a loss of all AC power. References 2 through 5 provide detailed analyses of these topics. Based on risk studies performed to date, the respits indicate that estimated total core melt frequencies from station blackout vary considerably for different plants and could be a significant. risk contributor for some plants. In order to reduce this risk, action should be taken to resolve the safety concern steming from station blackout. The issue is of concern for both PWRs and BWRs.

This guide primarily addresses the following three areas: (1) maintaining highly reliable AC electric power systems, (2) developing procedures and training to restore offsite and onsite emergency AC power should either one -

or both become unavailable, and (3) ensuring that plants can cope with a station blackout for some period of time based on the probability of occurrence of a station blackout at a site as well as the capability of restoring AC power in a timely fashion for that site.

One factor that affects AC power system reliability is the vulnerability to

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comen cause failures associated with design, operational, and environmental factors. Comon cause failure vulnerabilities are plant specific; therefore, generic guidance to improve the reliability of AC power systems by eliminat-ing specific single-point vulnerabilities is not practical. Existing W

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standards and regulatory guides include specific design criteria and guidance on the independence of preferred (offsite) power circuits (see General Design Criterion 17, " Electric Power Systems." and Section 5.1.3 of Reference 6) and the independence of and limiting interactions between diesel generator units at a nuclear station (see General Design Criterion 17, Regulatory Guide 1.6, " Independence Between Redundant Standby (Onsite) Power

. Sources and Between Their Distribution Systems," Regulatory Guide 1.75,

" Physical Independence of Electric Systems," and Reference :7). In developing the recomendations in this guide, the Staff has assumed that, by adhering to such standards, licensees have minimized, to the extent practical, single-point vulnerabilities in design and operation' that could resul' t ih a loss of all offsite power or all onsite emergency AC power.

Adoption of the proposed Section 50.63 and proposed paragraph (e) of GDC 17 would require all licensees and applicants to assess the capability of their plants to maintain adequate core cooling and containment integrity during a station blackout and to have procedures to cope with such an event. This guide presents a method acceptable to the NRC staff for determining the specified duration for whien a plant must be able to withstand a station blackout in accordance with the proposed amendment to GDC 17. Application of this method would result in selection of a minimum acceptable station blackout duration capability of either 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months depending on a -

comparison of the plant characteristics with those factors that have been - -

identified as significantly affecting the risk from station blackout. These -

factors include redundancy of the onsite emergency AC power system (i.e.,

number of diesel generators available for decay heat removal minus the number needed for decay heat removal), reliability of onsite emergency AC power

! sources (e.g., diesel generators), frequency of loss of offsite power, and probable time to restore offsite power. .

Licensees may propose durations different from those specified in this guide. The basis for alternative durations would be predicated on plant-specific factors relating to the reliability of AC power systems such as those discussed in Reference 2.

^

l

l i

l C. REGULATORY POSITION

1. Onsite Emergency AC Pcwer Sources 1.1 Reliability Program The reliable operation of the onsite emergency AC power sources should be ensured by a reliability program designed to monitor and maintain the .

reliability of each power source over time at a specified acceptable level and to improve the reliability if that level is not achieved. Guidance on acceptable emergency diesel generator reliability is given in Section C.1.2.

The reliability program should include surveillance testing, a reliability -

4 performance specification and a maintenance program. Surveillance testing should monitor performance so that if the actual performance falls below specified levels, corrective actions can be taken. Reference 8 provides an example of a performance technical specification to support such a program.

1.2 Maximum Failure Rate The maximum emergency diesel generator failure rate should be maintained at 0.05 failure per demand for all plants. For plants having an emergency AC power system redundancy as specified in group C of Table 2, the emergency diesel generator failure rate should be maintained at 0.025 failure per

, demand or less. . - -

1.3 Procedures for Restoring Emergency AC Power Guidelines and procedures for actions to restore emergency AC power when the emergency AC power system is unavailable should be integrated with plant-specific technical guidelines and emergency operating procedures developed using the emergency operating procedure upgrade program established in response to Supplement 1 to NUREG-0737, " Requirements for Emergency Response Capability" (Generic Letter No. 82-33).*

Modifications or additions to generic technical guidelines that are necessary to deal with a station blackout for the specific plant design should be identified as deviations in the plant-specific technical guidelines as recuired by Supalement 1 to NUREG-0737, and outlined in -

NUREG-0899, "Guicelines for t1e Preparation of Emergency Operating ._

Procedures."

6-

2. Offsite Power Procedures should include all actions necessary to restore offsite power and use nearby power sources

when offsite power is unavailable. As a minimum, the following potential causes for loss of offsite power should be considered: ,

a Grid undervoltage and collapse.

Weather-induced powe.- loss.

Preferred power distribution system faults ** that could result in the loss of normal power to essential'switchgear buses.

3. Ability to Cope with a Station Blackout The ability to cope with a station blackout for a certain time provides additional defense in depth should both offsite and onsite emergency AC power.

^

systems fail concurrently. Each nuclear power plant has the capability to remove decay heat without AC power for a limited period of time. Section C.3.1 provides guidance for determining the length of time that a plant is able to cope with a station blackout. Section C.3.2 provides a method to determine an acceptable minimum time that a pitnt should be able to cope with

, -a station blackout based on the probability of a station blackout at the site - -

as well as the AC power restoration capability for that site. If the plant's '-

. station blackout capability is significantly less than the acceptable minimum -

This includes items such as nearby or onsite gas turbine ~ge'neFators, portable generators, hydro generators, and black-start fossil power plants.

- Includes failures such as the distribution system hardware, switching and maintenance errors and lightning-induced faults.

. pm

duration, modifications may be necessary to extend the plants ability to cope with a station blackout. Should plant modifications be necessary, Section C.3.3 provides guidance on making such modifications. Whether or not modifications are necessary, procedures and training for station blackout events should be provided according to the guidance in Section C.3.4.

3.1 Evaluation of Plant-Specific Station Blackout Capability Each nuclear power plant should be evaluated to determine its capability to withstand and recover from a station blackout. The following considerations should be included'to determine the lengt'h o'f time the plant is able to cope with a station blackout:

1. The evaluation should be performed assuming tha,t the plant is operating at full power immediately before the postulated station blackout. -

2. The capability of all systems and components necessary to provide core cooling and decay heat removal following a station blackout should be identified, including DC battery capacity, condensate storage tank capacity, compressed air capacity, and instrumentation and control requirements. -

3. The ability to maintain adequate reactor coolant system: inventory to ensure that the core is cooled should be evaluated taking into consideration shrinkage, leakage from pump seals. and,in_ventory loss from letdown or other normally open lines dependent on AC power for isolation.

4. The design adequacy and capability of equipment needed to function in environmental conditions associated with a station blackout should be evaluated. All AC-independent decay heat removal systems 6

-i and associated equipment needed to function during a station blackout should. meet design and performance standards that ensure adequate reliability and operability in extreme environments, including hazards due to severe weather, that may be associated with a station blackout.- Work that has already been performed need not be duplicated. For example, if safety-related equipment needed during a total loss of AC power has been qualified to operate -

. during environmental conditions associated with a station blackout (e.g., without heating, ventilating, and air conditioning systems

operating), additional analyses need not be performed. ,

5. Consideration should be given to using available non-safety-related equipment, as well as safety-related equipment, to cope with a station blackout provided such equipment meets .the recommendations of item 4 in Section C.3.1 of this regulatory guide. In general, equipment required to cope with a station blackout during the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be available onsite. For equipment not located onsite, consideration should be given to its availability and accessibility in the time required including consideration of weather condition likely to pravail during a loss of offsite power.

6. Consideration should be given to timely operator actions that would

-- increase the length of time that the plant can cope with a station - "

blackout. For example, if station battery capacity is a limiting -

factor in coping with a station blackout, shedding nonessential loads on the DC batteries could extend the time until the battery is depleted. If load shedding or other operator actions are considered, corresponding procedures should be incorporated in the plant-specific technical guidelines and emergency operating procedures.

7. The requirements and recommendations contained in Reference 9 are considered to provide acceptable methods for performing this assessment.

ee

.-.. - ~ .. - ... .. . ~ . --- .- .~ -.

. a

-9

. . - ~ -

3.2 Hinimum Acceptable Station Blackout Duration Capability Each nuclear power plant should be able to withstand and recover from a station blackout lasting a specified minimum duration. The specified duration of station blackout should be based on the following factors:

The redundancy o'f the onsite emergency AC power system (i.e., -

number of power sources available minus the number needed fori decay heat removal), '

The reliability of each of the onsite emergency AC power sources (e.g., diesel generator),

The expected frequency of loss of offsite power, and The probable time needed to restore offsite power. -

A method to determine an acceptable minimum station blackout duration capability based on these factors is presented below. Table 1 presents the acceptable station blackout duration capability as a function of various specific site- and plant-related characteristics. Tables 2 and 3 provide descriptions of the emergency AC power configuration groups and the offsite power design characteristic groups, respectively, used in Table 1.' Table'2 ~

identifies different levels of redundancy of the onsite emergency AC power system. Table 3 provides a method to determine the offsite power design characteristic groups as a function of expected frequency of severe weather events, switchyard designs, number of offsite power circuits, and ability to restore offsite power. After identifying the appropriate groups from Tables 2 and 3 and the reliability level of the onsite emergency AC power t

4 l

== . :.:. . ..x ---

.=- . , . . - ==- -

.a.-.= =.. . . .. _ w_

1

! 1 sources (determined in accordance with Section C.1.1 of this regulatory guide) Table 1 can be used to determine the acceptable minimum station j ' blackout duration capability (4 or 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) for each plant.

! 3.3 Modifications to Cope with Station Blackcut . _,

. If the plant's station blackout capability, as determined according to the l

guidance in Section C.3.1 of this regulatory guide, is significantly less i than the reconnendr! plant-specific station blackout duration as developed according to Section C.3.2 of,this regulatory guide (or as justified by the licensee or applicant on some other basis and accepted by the staff)',

modifications to the plant may be necessary to extend the time the plant is able to cope with a station blackout. If modifications are needed, the i following items should be considered: .

J I 1. If, after considering load shedding to extend the time until .

. battery depletion, battery capacity must be extended further to

! meet the station blackout duration recommendations of Section C.3.2

! of this regulatory guide it is considered acceptable either to add batteries or to add a charging system for the existing batteries that is independent of both the offsite and onsite emergency AC power systems such as a dedicated diesel generator. - -

i 2.- If the capacity of the condensate storage tank is not sufficient to -

i remove decay heat for the station blackout duration recommendations of Section C.3.2 of this regulatory guide, a system to resupply the 1 tank from an alternative water source is an acceptable means to increase'its capacity provided that any power source necessary to provide additional water is dependent of both the offsite and onsite emergency AC power systems. ;

i i -

._. . . ~ . _ _ _ , . - _ _ _ , . - . . _ . . - - . - _ _ _ .

, .. .. :_ ._ ,1 _ - -

.;. c a x:--

3. If a system is required for primary coolant' charging and makeup, reactor coolant pump seal cooling or injection, or decay heat removal specifically to meet the station blackout duration recommendations of Section C.3.2 of this regulatory guide, the

following criteria should be met: _ _ _

. The system should be capable of being actuated and . .

---"- controlled from the control room; -

If the system must operate within 10 minutes of a loss of a'11 AC power, it shotrid be capable of being a'ctuated automatically; and The requirenents and recommendations conta.ined in Reference 9 are considered acceptable for specifying the performance and 1

design of equipment.

4. A system or component added specifically to meet the station blackout duration recommendations of Section C.3.2 of this regulatory guide should have limiting conditions for operation and surveillance requirements in the technical specifications .

consistent with those of other equipment needed to remove ~

decay heat during a loss of all AC power.

5. If a system or component is added specifically to meet the station blackout duration recommendations of Section C.3.2 of this regulatory guide, failures of added components that may be vulnerable to internal or external hazards within the design basis (e.g., seismic events) should not result in secondary failures causing a loss of emergency AC power systems or a loss of other safety-related equipment.

, . _ m + y v h- - = "" -

i . .

3.3 Procedures and Training to Cope with Station Blackout Procedures

and training should include all operator actions necessary to cope with a station blackout for the duration determined according to Section C.3.1 or C.3.2 of this regulatory guide, whichever is longer, and to restore normal long-term core cooling / decay heat removal once AC power is restored.

D. IMPLEMENTATION _

The purpose of this section is to provide information"t'o applicants and*

licensees regarding the NRC staff's plans for using this regulatory guide.

The proposed guide has been released to encourage public participation in its development. Except in those cases in which the appl.icant or 1,1censee proposes an acceptable alternative method for complying with specified portions of the Comission's reguhtions, the method described in the active guide reflecting public comments will be used in the evaluation of submittals by applicants for construction permits and operating licenses and by licensees who are required to ccmply with the proposed 50.63, " Station Blackout" and the proposed amendment to General Design Criterion 17 of 10CFR Part 50.

~

Procedures should be integrated with plant-specific technicaT gu~iifelinei ~

and emergency operating procedures developed using the emergency operating procedure upgrade program established in response to Supplement 1 of NUREG-0737. The task analysis portion of the emergency operating procedure upgrade program should include an analysis of instrumentation adequacy during a station blackout. '

l

E. REFERENCES

1. WASH-1400, " Reactor Safety Study," October, 1975.

2. NUREG-1032, " Evaluation of Station Blackout Accidents at Nuclear Power Plants, Technical Findings Related to Unresolved Safety Issue A-44," to

. .be published. . .

3. NUREG/CR-3992, " Collection and Evaluation of Complete and Partial Losses of Offsite Power at Nuclear Power Plants," to be published.

4. NUREG/CR-2989, " Reliability of Emergency AC Power Sources at Nuclear Power Plants," July 1983.

5. NUREG/CR-3226, " Station Blackout Accident Analyses (Part of NRC Task Action Plan A-44)," May 1983.

6. IEEE Std. 765-1983, "IEEE Standard for Preferred Power Supply for Nuclear Power Generating Stations."

7. IEEE Std. 387-1984, "IEEE Standard for Diesel-Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations."

8.,, Generic Letter 84-15, " Proposed Staff Actions to Improve _and Maintain Diesel Generator Reliability," July 2, 1984.

~ ~~

9. ANS-58.12. " Criteria for Evaluation of Response Capability for Loss of All AC Power (Station Blackout) at Light Water Reactor Nuclear Power Plants," Consolidated Draft Revision 2, May 21, 1984.

O em

Table 1. Acceptable Station Blackout Duration Capability (hours)I 2

_Emargency AC Power Configuration Group _

A B C

~

3 Maximum EDG Failure Rate Per Demand Offsite Power Design Characte'ristics 4 0.05 0.025' O.05 0.025 fi i i isi i i ll < / / /g i i i i r, I i .

/ , #

P1 y 4 1 4 1 4 4 I. /

r. . : i h---

I

' u'''z'

/

'I/ I l

P2 e

/

4 l 4 /

/ {8 l 8 % !

/ I / I JlI

/ I / b i q

,. 4----j - . _ _ _ . _ _ _ _ , . _ _ _ _

/ t i

' i J !

P3 4 4 / f 8 8 g

/ < 4 I ,!

I /

/ l .

l

. 's / / s /s A s i s s i s i ?* _T h'~ '%Yl _ _ _

I Yariations from these times will be considered by the staff if justification, including a cost-benefit analysis, is provided by the licensee. The methodology and sensitivity studies presented in NUREG-1032 (Ref. 2) are acceptable for use in this justification.

l 2

See Table 2 to determine emergency AC power configuration group.

3 Number of emergency diesel generator (EDG) failures in the last 100 valid demands divided by 100.

~

4 See Table 3 to determine categories P1, P2, and P3.

' ~

Tible*2. y AC'(EAC) Pcwer Configuration Groups Emergenc' Emergency AC pcwer Number of EAC power Number of EAC power sources required configuration grcup sources available l to operate AC pcwered decay heat removal systems 2 A 3 1 4 1 5 2 7___.

B 2 3 . .

j 4 2 4

. C 2 1 3 2 4 3 5 3 I

If any of the EAC power sources are shared between units at a multi-unit site, this is the total nucber of shared and dedicated sources for those units at the site. ~ ~ ~ ~

2 This number is based on all the AC loads required to remove decay heat (including AC-powered decay heat removal systems) to achieve and maintain hot shutdown at all units at the site with offsite power unavailable.

- For nonshared EAC power sources. - -- -- - - - - - ~~

For shared EAC power sources in which each diesel generator is capable of providing AC power to both units at a site concurrently.-

l M

,... .- . . .- .. . . ~ . ._ .. . -- . -...- - - . .

c 16 -

Table 3. Offsite Power Design Characteristic Groups i Group Offsite Power Design Characteristic

, Sites which have either one of the fol. lowing offsite power.

ifesigns:

- Sites with all offsite power sources connected to the plant through two or more switchyards or separate incoming transmission lines, with at least one of the AC sources electrically independent of the others, or P1 - Sites with all offsite power sources connected to the plant through one switchyard or through two or more switchyards l '

that are electrically connected, and if the normal AC power source is lost, there is an automatic transfer to an 4

alternate offsite power source. If this source fails also, -

there is one or more automatic or manual transfers of power j to another source of offsite power.

and

Sites with frequency of loss of offsite power due to extremely j severe weather less than 1 per 350 site years.*

and .

Sites which have one or both of the following characteristics:

- The capability and procedures to recover offsite (non-emergency) AC power to the site within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following a loss of offsite power due to severe weather, or r

- A frequency of loss of offsite power due to extremely severe weather less than 1 per 100 site years.**

~ ~ ~ ^ ~~

Sites with frequency of' loss of offsite power due to extremely -

i .

severe weather less than 1 per 100 site years *, and any one or .'

more of the'following three characteristics: ,

~

1. Sites with frequency of loss of offsite power due to - :

extremely severe weather greater than 1 per 350 site years *i_.

[:-

l or

- P2 2. Sites with all offsite power s'ources connected to the

. plant through one switchyard or through two or more switch-yards that are electrically connected, and if the normal ,

source of AC power is lost, there are no automatic transfers and one or more manual transfers (or alternatel

' one automatic transfer but no manual transfers)y, there is' to preferred

-or alternate offsite power 1 sources.

~

or 3.Siteswithfrequenc9'oflossofoffsitepowerdueto

^

severe weather greater than 1 per 100 site years **; without the capability and procedures to recover offsite (non-emergency) AC power to the site within 2-hours following a loss of offsite : wer due to severe weather.

P3 Sites with frequency of loss of offsite power due to extremely '

severe weather greater than 1,per 100 site years *. ._

m ' i, e

- , - , 4-- ~.-e . , - . . - , ,. ..-.- -,,., m. ..c.-, - - . . , . ,ww,.-muw.c.&

i l

Footnotes for Table 3

The estimated frequency of loss of offsite power due to extremely severe weather is determined by the annual expectation of storms at the site with wind velocities greater than or equal to 125 mph.

- The estimated frequency of loss of offsite power due to severe weather, f, is determined by the following equation:

, f = (1.8 x 10-4) h1+(27)h2 + (2.6 x 10-2)3 h

. p where by = annual expectation of snowfall for the site, in inches, h2 = annual expectation of tornadoes per square mile at the site, n3 = annual expectation of stonns at the site, with wind ,

velocities between 75 and 124 mph, and

~

The annual expectation of snowfall, tornadoes and storms are obtained from National Weather Service data from the weather station nearest to the plant or an interpolation, if appropriate, between nearby weather stations.

The basis for the empirical equation for the frequency of loss of offsite power due to severe weather, f, is given in Reference 2, Appendix A.

M

-w,-% --

G e ENCLOSURE 4 NUREG-1032 l

l l

f I 1

l I

[

i !

i/so/ss

~

DRAU] .

NUREG-1032 FOR C0fetENT EVALUATION OF STATION BLACKOUT ACCIDENTS AT NUCLEAR POWER PLANTS TECHNICAL FINDIf1GS RELATED TO UNRESOLVED SAFETY ISSUE A-44

+

e U. S. NUCLEAR REGULATORY COMMISSI0ft 0FFICE OF NUCLEAR REGULATORY RESEARCH l

e -t m -w + r - -1re-- , a- ewF- y --- * .- %-yaeyy~-y y -- e

TABLE OF CONTENTS Pace i

1 List of Figures. . . . . . . . . . . . . . . . . . . . . . . . . . . . .

List of Tables . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

iii

1. Executive Summary . . . . . . . . . . . . . . . . . . .:. . , :. c . .

1

2. Introduction and Technical Approach . . . . . . . . . . . . . . . . 5

~- :

3. Loss of Offs'ite Power Frequency and Duration . . . . . . . . . . . . 9

~

4. Reliability of Emergency AC Power Supplies. . . . . . . . . . . . . .. 17 T

~

5. Station Blackout Frequency and Duration .............. 37 O *

6. Ability to Cope with a Station Blackcut .............. 42

7. Accident Sequence Analyses . . . . . . . . . . . . . . . . . ._.:. .-

51

7- 8.- Evaluation of Dominant Station Blackout Accident Characteristics . .- .

69 -

F~ 9. Relationship of Other Safety Issues to Station Blackout . . ..-. . . -80 .

3 Appendix A. Development of Loss of Offsite Power Frequency and Duration .

Relationships Appendix B. Emergency AC Power Reliability and Station Blackout Frequency -

Modeling and Analysis Results Appendix C. Station Blackout Core Damage Likelihood and Risk

1 l

1 List of Figures .

Figure Page 3.1 Example Offsite Power System Diagram 3.2 Loss of Offsite Power Frequency vs. Duration -

3.3 Estimated Loss of,0ffsite Power Frequency and.

Duration for Representative Clusters 4.1 Simplified 1-of-2 Onsite AC Power Distribution Sys_ tem . .

4.2 Onsite Power System Functional Block Diagram -

4.3 Emergency Diesel Generator Demand Failure Probability -

Histograms for Years 1976-1982 4.4 Failure Contribution by Diesel Generator Subsystem 4.5 Onsite AC Sytem Unavailability for 18 Plants Studied .,

4.6 Percentage of Emergency Diesel Generator Failures -

Repaired vs. Time Since Failure 4.7 Generic Emergency AC Power Unavailability as a Function of EDG Reliability 4.8 Emergency Diesel Generator Failure to Run 5.1 Sensitivity of Estimated Station Blackout Duration Frequency for Several Representative Offsite Power. ,

Clusters

.:- 5.2 Sensitivity of Estimated Station Blackout Duration .;

Frequency for Several EDG Reliability Levels 5.3 Sensitivity of Estimated Station Blackout Duration . . _. .. ._.

Frequency for Several Emergency AC Power Configurations .. ._, .

7.1 Generic PWR Event Tree for Station Blackout 7.2 Generic BWR Event Tree for Station Blackout (BWR2-BWR3) 7.3 Generic BWR Event Tree for Station Blackout (BWR3-BWR6) 1

List of Figures (continued)

Figure Page ,

7.4 Time to Core Uncovery as a Function of Time at Which Turbine-Driven Auxiliary Feedwater Train - -

Fails 7.5 PWR Station Blackout Accident Sequence - : , -- -- : .- - --

5-7.6 BWR Station Blackout Accident Sequence ; -----

: - - - .:: :) -

Sensitivity of Estimated Station Blackout-Core 8.1 -----

Damage Frequency to Offsite Power Cluster, AC-Independent Decay Heat Removal Reliability, .

and Station Blackout Coping Capability 8.2 Sensitivity of Estimated Station Blackout-Core ,

Damage Frequency to EDG Reliability AC-Independent "

Decay Heat Removal Reliability, and Station Blackout -

Coping Capability 8.3 . Sensitivity of Estimated Station Blackout-Core --

Damage Frequency to Emergency AC Power Configurations, AC-Independent Decay Heat Removal Reliability, and Station Blackout Coping Capability --

8.4 Sensitivity of Estimated Station Blackout-Core . __- .- .-

Damage Frequency to Reducing EDG Common Cause ,

-3,3 :_- - -

Failure Susceptibility, EDG Reliability, -

.: :3:--

and Station Blackout Coping Capability 3-- - - -- 3 :- - :: ---

8.5 Estimate Core Damage Frequency Showing Uncertainty .. ..

Range for Four Reference Plants from NUREG/CR-3226 l

I ii

List of Tables Table Page ,

1.1 Summary of Station Blackout Program Technical Results ..

3.1 Sunnary of Total Losses of Offsite Power Data

- -:: re -

From U.S. Nuclear Power Plant- Sites from -

1968-1983 5- :- --

i :

=~~ ~ -" - - ---

4.1 Summary of Data from Diesel Generator Reliability Questionnaires .

4.2 Results of Onsite Power System Reliability Analysis 6.1 Station Blackout Effects on Plant Decay Heat Removal Functions >

6.2 Possible Factors Limiting the Ability.to Cope with a '

' =' - ~ ~ ~~ -

Station Blackout 7.1 Estimated Time to Uncover Core for Station Blackout Sequences with Initial Failure of AC-Independent

- '-' ~~

-"-~~ : :i Decay Heat Removal Systems and/or Reactor Coolant Leaks 7.2 Summary of Potentially Dominant Core Damage Accident "

Sequences :+ . :- .. :: :n 7.3 Containment Failure Insights : Err- i: : -

-e: : -

Containment Fission Product Release Categories and :- -

E:1 7.4 - - ::5::-

Failure Mode Probabilities for Station Blackout v - I" F 1: - -

Sequences :.- - :-_.

8.2 Sensitivity of Estimated Station Blackout Damage '- -~ - -

+- '

Frequency Reduction for Station Blackout Accidents with Reactor Coolant Pump Seal Failure Delay from 2 to 4 Hours and 4 to 8 Hours 9.1 Coupling Between External (and Internal) Events and Potential Plant Failures 111

1.0 Executive Sumary The complete loss of alternating current (AC) electrical power to the ,

essential and nonessential switchgear buses in.a nuclear power plant is referred to as a " Station Blackout." Because many safety systems required for reactor core decay heat removal and containment heat removal are -

dependent on AC power, the consequences of a station blackout could be severe. .

Existing regulations do not require explicitly that nuclear power plants be capable of withstanding a station blackout.

In 1975, the results of the Reactor Safety Study (Reference 1) showed that station blackout could be an important contributor to the total risk from nuclear pcwer plant' accidents. In addition, as operating experience has accumulated, the concern has arisen that the reliability of both the onsite and offsite emergency AC power systems might be less than originally anticipated. ,-

The Comission designated the issue of station blackout as an Unresolved Safety Issue (USI); a Task Action Plan (TAP A-44) was issued in July 1980, and work was initiated.to determine whether additional safety requirements were needed.

Technical studies performed in support of the program to resolve this safety issue have identified the dominant factors affecting the likelihood of station blackout accidents at nuclear power plants. A summary of the principal probabilistic results is provided in Table 1.1. These results are based on operating experience data; the results of several plant-specific probabilistic safety studies; and reliability, accident sequence, and consequence analyses performed in support of this unresolved safety issue.

In sumary, the results show the following important characteristics of station blackout accidents.

(1) The variability of estimated station blackout likelihood is potentially large, ranging from approximately 10-5 to 10-3 per reactor-year. A

" typical" estimated frequency is on the order of 10-4 per reactor-year.

i 1 l (

l NUREG-1032 j 1

Table 1.1. Summary of Station Blackcut Program Technical Results Operational Experience Loss of offsite power (per year)

. Average 0. 1- ,

m Range 0-0.4 .

Time to restore offsite power (hours) -

Median 0.5

, 90% restored 3.0 Emergency diesel generator reliability per demand ?

ii Average 0.98 Range -

0.9-1.0 Median emergency diesel generator repair time (hours) 8 i

l Analytical Results .-

Estimated emergency AC power systems unavailability range _ r u :,-- . :: e - ; - -

(per demand) 10~4 --

10f2. .

Estimated station blackout frequency range (per year) 10-5 -

10-3 .

l Estimated station blackout - core damage frequency range (per year) 10-6 -

10-4 NUREG-1032 - . _ . . - _ _. -

(2) The capability of restoring offsite power in a timely manner (less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) can have a significant effect on accident consequences.

(3) The onsite AC power system redundancy and individual power supply reliability have the largest influence on station blackout accident likelihood. -

u (4) The capability of the decay heat removal system to cope with long duration blackouts (greater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) can be a dominant factor influencing the likelihood of core damage or core melt for the accident sequence. -

(5) The estimated frequency of station blackout events resulting in core damage or core' melt can range from approximately 10-6 to greater than 10-4 per reactor-year. A " typical" core damage frequency estimate is on .,

the order of 10-5 per reactor-year. .

(6) 'Information currently available indicates that containment failure by overpressure may follow a station-blackout-induced core melt. Smaller, low design pressure containments are most susceptible to early failure (possibly less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ). Some large, high design pressure contain-ments may not fail by overpressure, or the failure time could be on the order of a day or more. _. . .

It was found that losses of offsite power could be cat.egor.ized .as .th.os_e _ . - . _ : :,-

resulting from plant-centered faults, utility grid blackouts, or severe weather-induced failures of offsite power sources. The industry average frequency of total losses of offsite power was determined to be about 0.1 per site-year, and the median restoration time was about one-half hour. The factors which were identified as affecting the frequency and duration of offsite power losses are:

l (1) Design of preferred power distribution system, particularly the number and independence of offsite power circuits from the point where they I

(

enter the site up to the safety buses; l 1

I NUREG-1032 l

l (2) Operations which can compromise redundancy or independence of multiple offsite power sources, including human error; (3) Grid reliability and security, and the ability to restore power to a nuclear plant site with a grid blackout; and l' : (4) The hazard from, and susceptibility to, severe weather conditions which ...-

- can cause loss of offsite power for extended periods.:: ::.:- --e -- "-e- :- ::.

~

~ A design and operating experience review, combined with a- reliability- - -

~ analysis of the onsite, emergency AC power system, has shown that there are a variety of potentially inportant failure causes. The typical unavailability of a two-division emergency AC power system is about 10-3 per demand, and the typical individual emergency diesel generator failure rate is about 2 x .

10-2 per demand. The factors which were identified as affecting the -

~ ~~~~ emergency AC power system reliability during a loss of-offsite power are: - -

(1) Power supply configuration redundancy; -

(2) Reliability of each power supply; (3) Dependence of the emergency AC power system on supporteor auxiliary -

n : -

cooling systems, control systems; and ---

a. - -

7

- '(4) Vulnerability to common cause failures associated with design,

~

operational and environmental factors. -3 - .- :

The likelihood of a station blackout progressing to core damage or core melt is dependent on the reliability and capability of decay heat removal systems which are not dependent on AC power. If sufficient capability exists, additional time will be available to permit an adequate opportunity to restore AC power to the many systems normally used to cool the core and remove decay heat. The most important factors involving decay heat removal during a station blackout are:

NUREG-1032 I

1 (1) The starting reliability of systems required to remove decay heat and maintain reactor coolant inventory; (2) The capacity and functionability of decay heat removal systems and auxiliary or support systems which must remain functional during a station blackout (e.g., DC power, condensate storage); -

(3) For PWRs, and BWRs without reactor coolant makeup capability 3during i" ~" :~ +' -

station blackout, the magnitude of reactor coolant pump seal leakage;

~ ~ ~ ~

- ~

and (4) For BWRs which remove decay heat to the suppression pool, the ability f to maintain suppression pool integrity and operate heat removal systems :

t at high pool temperatures during recirculation. j

~

It was' determined by reviewing design, operational, and location factors,

~

that the expected core damage frequency from station blackout could be

~

maintained around 10-5 per reactor-year or lower for all p1 ants. The ability ~~~

'~

to cope with station blackout durations of at least 4 and perhaps 8 to 16 [

hours and emergency diesel generator reliabilities of 0.95 per demand or better with relatively low susceptibility to common cause failures would be necessary to reach this core melt likelihood level. +: ':c ~~ '~~ i"~ ~

7, ;

2. Introduction and Technical Aporoach

~~

The term " station blackout" refers to the complete loss of~ alternating :i current electrical power to the essential and nonessential buses in a ;

nuclear power plant. Station blackout, therefore, involves the loss of {

offsite power concurrent with the failure of the onsite emergency AC power system. Because many safety systems required for reactor core cooling and decay heat removal and containment heat removal are dependent on AC power, !

the consequences of station blackout could be severe. ;

I The concern about station blackout arose because of the accumulated !

experience regarding the reliability of AC power supplies. A number of operating plants have experienced a total loss of offsite electrical power, 4 NUREG-1032 I

and more occurrences are expected in the, future. During these loss-of-offsite-power events, the onsite emergency AC power supplies were available to supply the power needed by vital safety equipment. However, in ,

some instances, one of the redundant emergency power supplies has been unavailable, and in a.few cases there has been a complete loss of AC power. ;

l-* ~ ~ During these events, AC power was restored in a short time without any

serious consequences. In addition, there have been numerous instances where t --

~52:'~ 5 E emergency diesel generators failed to start and run in response to testsv- -

conducted at operating plants. :; : 5:

--~ .

t

' Results of the Reactor Safety Study (Reference 1) showe.1 that for one of the two plants evaluated, a station blackout accident could be an important contributor to the total risk from nuclear power plant accidents. Although a this total risk was found to be small, the relative importance of the .

station blackout accident was established. This finding and the accumulated "

I ~ di~e sel generator failure experience increased the concern about station.

! blackout.

An analysis of the risk from station blackout invo'lves an assessment of the

> likelihood and duration of the loss of offsite power, the reliability of onsite AC power systems, and the potential for severe accident sequences --

- - after a loss of all AC power. These topics were investigated under _ . ::. 33 -

~

I' Unresolved Safety Issue Task Action Plan (TAP) A-44. This plan . included the . -- :

following major tasks: .

(1) Estimating the frequency of station blackout at operating-U. S. nuclear. - -

power plants. This analysis consisted of two parts: .

L (a) Estimating the frequency of loss of offsite power for various plant locations, and f (b) Estimating the probability that the onsite AC power system will

fail to supply AC power for core cooling.

i f

NUREG-1032 -

1 (2) Determining plant responses to station blackout and the risk associated I with station blackout-initiated accident sequences. The scope of this ,

4 investigation included:

(a) Reviewing the shutdown cooling system design and assessing its

' ~

=-!

capability and re. liability under prolonged station blackout conditions, and

~~

C (b) Reviewing the containment design and its ability to withstand temperature and pre'ssure buildup during a prolonged loss of-AC power.

~

The principal focus of this TAP was on the reliability of emergency AC power supplies. This approach was taken for several reasons. First, the questions raised about the reliability of onsite emergency power supplies basically were responsible for identifying station blackout as a safety issue. Second, .'

if safety improvements are required, it is easier to analyze, identify and- ,

~ implement them for the onsite AC power system than for the offsite- AC power-

~ -

supplies or for the AC-independent decay heat removal system. For example, offsite power reliability is dependent on a number of factors such as regional -

electrical grid stability, weather phenomena, and repair and restoration capability, which are difficult to analyze and to control. Also, the capability of a plant to withstand a station blackout depends on those - -

decay heat removal systems, components, instruments and controls which are :- - m-independent of AC power. These features vary from plant;to plantc requiring ~ - -

considerable effort to analyze all of them or to assure that the plants indeed have that capability. Third, significant progress has been made on improving operating PWRs by backfitting the auxiliary feedwater: system to -

make it independent of AC power. In addition, USI A-45, Shutdown Decay Heat Removal Requirements, is reviewing the adequacy of shutdown decay heat removal systems for nuclear power plants. Thus, the reliability of emergency AC power supplies is of principal importance to USI A-44.

A preliminary screening analysis was done to identify plants most likely to suffer core damage due to a loss of all AC power supplies. The intent of this work was to survey the frequency and implication of station blackout l

l NUREG-1032

[

accidents in operating plants and identify any especially high risk plants which might require further analysis or action on an urgent basis. Initial results showed no such plants.

Following this initial analysis, a more detailed evaluation of station blackout concerns was perfomed. Since this issue centers around a concern .

~ for the adequacy of AC. power supply reliability, typical offsite.and emergency

~ ~ ~AC power supplies were evaluated including a review of. past; operating . 3 + -:-:

~

fr (failure) experience. This effort was limited to power. supply availability - e--

and did not include an evaluation of power distribution adequacy or power capacity requirements.

Enfomation on loss of offsite power was collected from licensee event reports (LERs), responses to an NRC questionnaire, and various reports .,

prepared by the utilities. Most of the event descriptions contained in the ,

~

~ LERs and other documentation in the Nuclear Regulatory: Commission files did -

i not contain sufficient infomation to provide an accurate data base to estimate frequencies and durations of losses of offsite-power. For example, -

in one case a licensee reported offsite power restoration time to be 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , but actually one offsite power source was restored in 8 minutes, and all.

offrite power was restored in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Since restoration of one source-of 4

--":offsite power terminates a loss of offsite power, the-documented description .

~= was not-accurate enough. In some other cases, offsite: power was available to -;

~ be~ reconnected, but the plant operators did not reconnect it for.some time -: c

~a fter it~was available because onsite power was available. -Because of the -- --

need for more accurate data, additional infomation was obtained by working . -

! ' closely with the Institute of Electrical and Electronics Engineers (IEEE) and .

! the Electric Power Research Institute (EPRI) who contacted utility engineers to obtain better descriptions of the causes, sequences of events', and the times and methods of restoring offsite power.

In order to provide a consequence perspective, station blackoilt accident sequences and associated plant responses were analyzed. The Interim Reliability Evaluation Program was one source of information for developing the shutdown cooling reliability models and accident scenarios needed to perform this evaluation.

NUREG-1032 ;

i

1 This NUREG report sumarizes the results of the technical evaluations discussed above. Details of the technical assessments are reported in references 3, 4, and S. Additional technical evaluations provided in this NUREG were derived from thase references to coalesce that material and extend the analysis to obtain the broader insights and bases necessary to resolve this issue ~in'an integral manner considering plant differences. These supplemental

~

~~~2 analyses ~are described in Appendicies A, B, and C of this report. __

7' 3.- Lo~

s s' of Offsite Power Frequency and Duration . + _

e- -

The offsite or preferred power system at nuclear power plants consists of the following major ccmponents:

Two or more incoming power supplies from the grid; .

One or more switchyards to allow routing and distribution of pcwer ,

~~ -

within the plant;

<

One or more transformers to allow the reduction.of-voltage to levels

~

needed for safety and non-safety lystems within the plant;

The distribution systems from the transformers to the switchgear buses.

~ Figure 3.1 provides an example of an offsite power system design used for

~

nuclear power plants. During normal operation, AC power;is typically - - -- ---

- provided to the safety and non-safety buses from the main generator through --

~ ~ the auxiliary transformer, or it may be supplied directly through a start up -_

E transformer. A minimum of two preferred power supply circuits must be -

provided. Sources of offsite power other than the grid may-also be provided -

' as alternate or backup sources of power. These may include nearby (or onsite) gas turbine generators, fossil power plants, and hydroelectric power facilities. A loss of offsite power is said to occur when all sources of offsite power become unavailable causing safety buses to become deenergized and initiating an undervoltage signal. Some loss of offsite power transients will be very short--Just long enough to allow switching from one failed source to another available source. Because of the short duration characteristics of this type of loss of offsite power transient, it is not of NUREG-1032 , , - _

-.9

1 l

l n a a a ,

a a a a a ,

g ..

. g .

4 _ . _ _ . _

4 MMA MM MM MM MM AAAA AAAA

~

U 1r 1r1r 1r if NC NC NONSAFETY NO NONSAFETY NO WIN CLASS 1E CLASS 1E GENEnATOn CLA38IE CLASS 1E OtVISION I DIVISION 2 DIVISION I OlvlSION 2 e e .

I l 4 I

I g - - AuTOuAric TnANSFEn-----------t-------- J - --

L _ _ _ _ _^u,T_myLC LnA,NSJEn _ _ _ _ _ ; .,_ _ _ _ - __ _,

I l~ Figure 3.1. Example Offiste Power System Diagram .

1 I

l NUREG-1032 .

concern relative to station blackout. This type of loss of offsite power transient is better described as an interruption. If switching errors or failures of alternate sources of power compound the si'.uation and longer tenn repair, restoration, or alternate power source actuation activities are required, the loss of offsite power transient can be of significance. This type of loss of offsite power event is referred to as a total loss of offsite -

power.

Although total loss of offsite power is a relatively infrequent occurrence at -

nuclear power plants, it has happened a number of times in the past, and a data base of information has been compiled from past experience (Ref. 2, 3).

Historically, a loss of offsite power occurs with a frequency of about once per ten site-years. The typical duration of these events is on the order of one-half hour. However, there has been experience at some power plants in -

which the frequency of offsite power loss has been substantially in excess of ,

the average, and in other instances the duration of offsite power outages has greatly exceeded the norm. Table 3.1 provides a summary of the data on the total loss of offsite power events through 1983. -

Since design characteristics, operational features, and the location of _

nuclear power plants within different grids and meteorological areas can have- - --

a significant effect on the likelihood and duration of loss of offsite power events, it was necessary to analyze the generic data in more detail. The data have been categorized into plant-centered events and area or weather-related events. Plant-centered events are those in which the design and operational characteristics of the plant itself play a role in the likelihood of the loss of offsite power. Area or weather events include the reliability -

of the grid or external influences on the grid which have an effect on the likelihood and duration of the loss of offsite power. The data show that plant-centered events account for the majority of the loss of offsite power occurrences. The area blackout and weather-related events, although of lesser frequency, typically account for the longer duration outages with storms being the major factor. Figure 3.2 provides a plot of the frequency and duration of loss-of-offsite-power events due to plant-centered faults, grid blackout, and severe weather based on past experience at nuclear plant sites.

NUREG-1032 L

Table 3.1 Summary of Total Losses of Offsite Power Data for U. S. Nuclear Power Plants Sites from 1968-1983 .

Frequency of ,:

Causes of Loss of Occurrence . :- Med.ian

'~ Offsite Power Number (per site-year) e - - - Duration _-

Plant-Centered 30 0.056 0.3. h rs Grid Blackout 10 0.019 0.7 hrs Severe Storm 6 0.011 2.6 hrs -

Total 46 0.086 0.5 hrs .

Number of site-years through December 1983: 533

~

(Data exclude sites with one offsite power connection, ;4umboldt Bay, Lacrosse, -

~

and Big Rock Point prior to March 1968, and eight events which have been: -

classified as " interruptions" of offsite power supplies).: -

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i i

I NUREG-1032 -

13 -

1 1

Appendix A to this report provides a more thorough discussion of the technical bases for the loss of offsite power frequency and duration characteristics discussed in the remainder of this section.

Plant-centered failures typically involve hardware failures, design deficiencies, human errors (maintenance and switching) and localized weather- -

induced faults (lightning and ice) or combinations of. thase failure types.

The frequency of plant-centered loss of offsite power was-not found-to be . - 3

~

~~ -

strongly correlated with any particular desijn factor. -However,-ae modest - .- ,- -

3

~

correlation has been observed between the duration of plant-centered loss-of-- -

offsite-power events and the independence and redundancy of offsite power circui:s at a site. In this regard it has been observed that a site with several immediate and delayed access circuits will generally recover offsite power more promptly than a site with the minimum requirements. However, virtually all plant-centered faults of relatively high frequency are -

recoverable within a few hours. .

~

plant location plays an important role in loss of offsite power events. --

Factors shown to be significant are (1) the reliability of the crid from which the nuclear pcwer plant draws its preferred power supply and (2) the likelihood of severe weather in a location which can cause damage to the grid

~ distribution system and hence a loss of power to the plant. Traditionally, past analyses have focused on grid reliability as a dominant. factor in - -

~

developing the likelihood relationship for the loss of offsite power at a -:

~~~

si~te'. However, a review of the historical data shows that approximately 15. -- ~

percent of all losses of offsite power have been caused by grid problems,

~' and, in fact, a large percentr.ge of grid-related losses of offsite pcwer can .

be traced to one utility's system. The historical grid reliability of that system dominates the data, distorting the perspective on the contribution of grid failure to loss of offsite power frequency. This finding should not be unexpected when one. recognizes that current distribution and dispatch systems are well coordinated. Utilities shed loads when possible and generally protect their grid from overloads and faults that could cause grid loss in NUREG-1032 . _ - . _. _-

l i the various day-to-day operations. Moreover, when there is a loss of power on the grid, the first activity that is usually undertaken is the restoration of power to the electric generation plants so that the grid may be restored ;

with appropriate power supplies. 'In fact, during the northeast blackout of 1965, power was restored to a nuclear power plant in New England within about

' ~' o'n'e-half an hour of the grid collapse, while power was ,not restored to the grid in general for as long as 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more.

~

: --:--- . 1. . ... .. . ..

3. .

3 y . .

'~ ~

With 'the exception of a few utility systems, large grid-disturbance's' are

- :-- =

[7'+

'~~~ ~

relatively' infrequent, and, again with few exceptions,- the duration of power outages at power plants due to grid disturbances is relatively short. An

+

identified systemic weakness is usually corrected as soon as practical. It is the unidentified weaknesses that result in grid failures. In the absence of an historical trend, previous operating experience related to grid .-

reliability is not necessarily an indication of. future problems. This is ,

unless a known pre-existing weakness has not been corrected. Because grids in the U.S. are generally very stable, and system planning activities are.

directed at maintaining and improving that stability, grid reliability is not '

necessarily the principal indicator of the likelihood of loss of offsite power.

Severe weather, such as local or area-wide storms, can disrupt incoming power

~ supplies to the plant. In fact, there have been a number of failures of l-

~

offsite power at nuclear power plants which were weather-related. These

'~~' '

~'an be divided into two failure groups. - - - -

c :e : - .

~

(1) Those for which the weather caused the event but: did:not affect the

~~'

time to restore power; and - =

(2) Those for which' the weather initiated the event and also caused conditions over a sufficiently broad area such that power was not or could not

' have been restored for a long time.

NUREG-1032 -

- , _ _ _ - - - ~ . .

l l

l i

i The former group includes lightning and most other weather events that are not too severe. They can cause a loss of offsite power, but their severity generally does not contribute in any significant way to long duration offsite ,

power losses. These types of weather-related offsite power outages have been treated as either plant-centered or grid losses of offsite power. The second group includes locses of offsite power due to major storms, hurricanes, high -

winds, snow and ice stonns, and tornadoes! The loss of offsite power

~ ~ ~ expectation frequency of this group is relatively small. On the other hand,

~ ~ ' ithe likelihood of restoring offsite power in a short time for- this group;.is -m : --

~a lso~ relatively .small . Whereas it is expected that dispatch and plant -

personnel actions can influence substantially the duration of area-wide grid disturbance outages causing a loss of offsite power at nuclear power plants, it is not expected that severe weather conditions can be significantly affected in terms of the expected duration of offsite losses at nuclear power plants.

Therefore, one would expect severe weather to dominate the restoration ,

characteristics for long duration cutages. The redundancy, separation, and independence of the offsite power system may affect the likelihood of some weather-related losses such as those induced by tornado strikes. The -

effectiveness of these design considerations on reducing the likelihood of other types of weather-related outages has not been shown.

~~'~~ 'lt'has been shown that there is a potentially large variation in the annual :-

expected frequency of offsite power outages for different nuclear power .

~ ~' plants depending on their design and location. Also, a variation has been -- :--

~ ~ ~ ~ observed in the duration of offsite power outages that have occurred at -- - --

various nuclear power plants. The expectation of long duration outages is dominated by the severe storm likelihood and to a lesser extent by grid ,

blackout likelihood and the ability to restcre power to the site during grid loss. Grid-related losses are only important when the frequency of occurrence greatly exceeds the national average.

Modeling and analyses have been performed to determine the relationship between design and location factors and loss of offsite power frequency and duration representative of most nuclear ~ power plant sites in the United NUREG-1032 .

States. This work is described in Appendix A. Figure 3.3 provides a plot of the expected frequency and duration for loss of offsite power for site, design, grid, and weather characteristics which have been found to " cluster" reasonably well. The factor which most predominantly affects the characteristic groupings is severe weather hazard. Table 3.2 provides a

~

' definition of the site characteristics which make up the loss of offsite -

' power clusters shown. -

iReliability of Emergency AC Power Supplies

'~

24, 3:---- :: ::

The~ emergency AC power system is designed to provide an alternate-or backup power supply to the offsite power sources. A simplified one line diagram of a typical emergency AC power system is provided in Figure 4.1 for reference.

In the event that the offsite power system is lost, an undervoltage condition will exist on the safety buses causing actuation of the emergency AC power _

system. The emergency AC power system provides sufficient functional-capability and redundancy of the power requirements for systems needed to mitigate the consequences of a design basis accident. This typically includes a requirement to actuate emergency AC power supplies and make them available for loading within about 10 seconds after receiving an actuation signal. The emergency AC power system also meets the single failure criterion

'when applied to design basis accidents.

~

~ ' Emergency AC power is generally provided by diesel generator systems, - -

=-

~~a lthough other sources such as gas turbine generators or hydroelectric power - -

r --

are used at some plants. Because of the preponderance of diesel generator .

uscge, that power supply type will be the principal focus of emergency AC power system discussions in this report. Figure 4.2 is provided for reference to identify the typical subsystems and support systems that are needed for successful operation of the emergency diesel generator.

NUREG-1032 - - .

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Figure 3.3. Estimated Loss of Offsite Power Frequency and Duration for Representative Clusters r

6 NUREG-1032

- 18 -

1 I .

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I

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Table 3.2 Characteristics of Some Loss of Offsite Power Clusters Which Affect Longer. Duration Outages

_ __ C1uster 2_ _ Offsite Power Design Group II, I2, or I3* and-- locatedi n very low L_ ._

. severe weather hazard area or susceptibility to s'ev~ere" weather =~~~ ~i;~.~5 2 hazards are very low -

- 2 -N "

_____Clustet_4_ . fioderate to high severe weather h zard area nd susceptibility . .- I. _.

~~ , ~ - - - - - _ . . _ . . . _ _ . ._ _ ._

~

Cluster 5 -'Very high severe weather hazard area and susceptibility __ _ - . .

Cluster 7 - Average grid reliability and combinations of offsite power system design and severe weather hazard / susceptibility which most closely

..... approximate the national average for nuclear plan _ts- _ _.

Q=.

~:~~~.

~~

- - ; . ~ ,. =. .

~ ~ ~ ~ -

See Appendix A for definitions of design groups II, I2, and I3.

)

i !

l NUREG-1032 i

MalN OFFSITE OFFSitt M AIN

UNif POWER UNIT POWE R ww m 'w wew a ww EMERGENCY CMCRC(NCY sus l' sus 2

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1 A ANSFORMCa Figure 4.1. Simp 11ffed 1-of-2 Onsite AC Pcwer Distribution System -

l l

NUREG-1032 i'

l' -

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-- -____________ ~ _

q CC ' \ EMERGENCY j

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Figure 4.2 Onsite Power System Functional Block Diagram 1 NUREG-1032 .

g a Emergency AC power systems typically consist of two diesel generators, any one of which is sufficient to meet AC power load requirements for a design basis accident. This configuration has been designated by its success ,

criterion: one out of two or more simply 1/2. Configurations exist where three or four or more diesel generators are used at single-unit sites, anl in

~

- some cases diesel generators are shared at multi-unit sites. These systems ,

also can be described by their success criteria, or number of diesel-

~+ : ': generators required per number provided. However, for the purpose of : :3 -

3- -

evaluating-the station blackout issue, the success criter.ia will- be . defined as - -

' the number of diesel generators required to maintain a stable core cooling --- - - -

and decay heat removal condition with all offsite power sources unavailable.

The emergency AC power configurations which exist in the United States have been identified as follows:

t

1. Emergency AC power supplies dedicated to one unit .

A. 1/2 B. 1/3 -

C. 1/4 D. 2/4

2. Emergency AC power supplies shared between two units . -

.- -=.

l A. 1/2 B. 2/3 .

C. 2/4 D. 2/5 ,

E. 3/5

3. Emergency AC power supplies shared between three units A. 3/8 l

l l

l l NUREG-1032 , _

A closer inspection of emergency AC power supply requirements may produce some variations on these configurations, however, they represent a wide variety in system success criteria for reliability evaluations.

Emergency AC power system design variability is further complicated by

$- dependencie3 on certain support systems which by themselves have a multitude- -

inf designs. These support systems include cooling systems :(air or: water), DC

'i ; ': power, and heating, ventilation and air conditioning (HVAC) systems. -In :t- .:+-:- :r "3 additidd,'~it is recognized that raintenance and testing activities : vary: -' :.- :! e

' considerably which can effect the reliability of the emergency AC power = +-

~' - "

system.

Emergency AC power systems can be considered in two separate parts: the

' power supplies and the power distribution system. In general it has been ,

found that the details of the emergency AC power distribution system from the -

safety (switchgear) buses to the safety components are not significant contributors to the unavailability of AC power for station blackout concerns.

This is because many independent, separate, and diverse distribution system components must fail to cause' loss of all AC power to the safety systems. It is recognized that fires and earthquakes have the potential to cause such distribution system failures. However, these hazards have been studied as separate safety issues, and as such, they were not the subject .of systematic --

2 -

assessment as part of the station blackout issue.

Substantial operating experience data were investigated to identify and estimate important reliability characteristics of emergency diesel generators (Ref. 4). Diesel generator reliability performance information was collected from 36 nuclear power plants with 90 diesel generators. A summary of the emergency diesel generator statistical data collected is provided in Table 4.1. In addition, information regarding diesel generator outages and downtime was obtained from responses to TMI Action Plan (Ref. 6) requests by plants with 58 diesel generators, and over 1,500 LERs covering a 5 year period from 1976 through 1980 were reviewed for failure information. Frem NUREG-1032 _

Table 4.1 Summary of Data from Diesel Generator Reliability Questionnaires No. of units - 45 No. of reactor-years - 194 No. of diesel generator-years - 450 Start Attempts and Failures for Tests and Actual Demands No. of Auto-Auto Start No. of Fail- Start Fail-No. of Fail- ures/ Fail- ures/ Unavail- Unavail-Category Demands ures Demand ures Demand able ability

" Test -13,665 253 0.019 55 0.004 - - - 0.006

-Loss of 100 5 0.05 3 0.03 3 0.03 -

offsite power

All 539 14 0.026 5 0.009 3 :- 0.006 emergency _. .

demands

- ^ -

Failure to run: 2.4 x 10-3/hr ,

(based on 314 attempts at scheduled run time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or more with 9 failures to run during these attempts)

Updated from data reported in NUREG/CR-2989.

NUREG-1032 . ., .. _ . . - . _ _ - - __ _. _ . _ . - .

l l

analysis of this operating experience it was found that on the average diesel generators failed to start, load, or continue running approximately two times out of every 100 demands. It was also observed that during the actual loss of offsite power events which occurred through 1983 there were 19 instances in which one or more diesel generators failed, operated in a degraded condition, or were otherwise unavailable. And, during the period 1976 through 1982, there .

. r were 45 instances of multiple diesel generator outages identified of which 11 ;.

were classified as common cause failures.

Figure 4.3 provides histograms of the probability of emergency diesel generator failure on demand for the years 1976 through 1982. As can be seen, although the average failure on demand likelihood is about 2 x 10-2 , there is a significant spread from the highest to the lowest demand failure rate. The average failure rate and range have not changed substantially during this l period. A review of the data has not identified any particular type of failure ]

-- as being the most dominant. At least in part, the reasons for this are that there are several different types of diesel generators having different support

and auxiliary system designs operating at nuclear power. plants, 6ad that ,

maintenance and test activities are not standardized within the nuclear l

industry. The percentage contribution of failure by subsystem is provided in Figure 4.4. In general, sufficient information was not available to add high

confidence to the correlation of root failure causes with specific designs and _

operational factors. The data indicate that approximately 80 percent of the failures are due to hardware-related arablems and 20 percent are due to human error.

This is not to imply that any one particular diesel generator will be . .

susceptible to all possible failure modes with equal importance. It is more likely that a few specific defects may exist, and if not discovered and corrected, future failures may occur. The failures observed can be classified into the following general types:

1. Design and hardware failures related to mechanical integrity or various failure modes in the diesel generator subsystem such as fuel, cooling, starting, and actuation; NUREG-1032 . ~ _ . , . . . . . . . . ..T.'......L.". . . . . . . . . . . . .

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Figure 4.3 Emergency Diesel Generator Demand Failure Probability Histograms for Years 1976-1982 i

l NUREG-1032

- 26 -

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hllllr-lll SUBSYSTEM Figure 4.4 Failure Contribution by Ofesel Generator Subsystem NUREG-1032 i

j 2. Operation and maintenance errors related to the correctness and adequacy

< of procedures or training, and human, factors including the potential for ,

errors of commission and omission; and

3. Failures which occur in support systems, or at interfaces with support

- and other systems, which can involve DC control power, service (or raw) ,-

water cooling, environmental control (air temperature and-quality), and interface with the normal AC power system. u 3: 2 .r-- -3 . . :; .

' MMultiple diesel generator failures can occur when a fault-or degradation: - -

~

exists involving a common factor or dependency for two or more diesel generators. They may also include design and operating deficiencies similar to those previously mentioned, but in this case degradation or failure occurs concurrently in multiple diesel units. ' For instance, a defective crankshaft -

design may be such that mechanical failure is highly likely to occur after a I certain amount of usage. If two or more diesel generators reach that usage - -

level at nearly the same time, concurrent failures may result. As another l

- -example, defective maintenance procedures and training could result-in human .

l errors causing failure or simultaneous outages of two or more diesel units.

i Another type of common cause failure is related to the existence of single

point vulnerabilities. Examples include a check valve in-a header of a . _

l cooling water supply, the unrecognized dependence on an obscure single -- - - -

=

r ~ contro1' circuit or element, and use of common fuel supplies' and containers., --- .-

f l Lastly, common cause failures can be related to coninonality of location with j regard to environmental conditions for which adequate protection is not- -

j provided. These can include items like fire, flood, dust, corrosive elements in air, or temperature and humidity extremes.

Consideration was given to the failure modes, causes, and failure rates derived from the operational data in assessing the reliability of emergency AC power systems. Reliability analyses performed by Oak Ridge National

.NUREG-1032 . - . - . - - . - - - . - - - . .- . , - . .

Laboratory (ORNL) for 18 nuclear power plant AC power configurations and the j plant-specific failure data were applied to derive typical system unavailability estimates. Figure 4.5 shows a histogram of the onsite AC power results for the .

l 18 plants studied. The results of this work, summarized in Table 4.2, shows the diesel generator configuration studied, the calculated range of unavailability on demand and the dominant failure causes for each group analyzed. Not ,

surprisingly, for the least redundant system configuration,,the independent diesel generator failure likelihood is the most dominant failure. .= factor. .As system redundancy is increased, connon cause failures become more important.

SConnon cause failures involving hardware , failure, human-error;,.and. dependent. . . , ., .

system failures were found to be important. _ .

, For the most part, power supply outages resulting from testing and maintenance were not found to be large contributors to system unavailability. However, ,

a few cases were identified in which extensive maintenance outages could

~

.cause significant system degradation in availability.. The. quality of test.and .

maintenance procedures, however, can be an important factor affecting system

)- reliability. Lower than average, human-error-related diesel ._ generator . . . .. . .

failures were observed when procedures were clearly written and had sufficient level of detail, including complete check lists for operations personnel to verify proper indication of nonnal values after maintenance.

~~ ~

-The impact of dependen.t systems, such as service water cooling and DC power,. . . .

- on'the reliability of the emergency AC power system is. plant-dependent. ,The .

ORNL analyses did not go into detail on the reliability of those support ,

systems. However, dependent system failures affecting the emergency.AC power system seem to be dominated by single-point passive failures or human. error, . . . _

and results do show that a highly unreliable , support system can cause a' highly i unreliable AC power system. Since these support and auxiliary systems also tend to be important for the operation of decay heat removal systems, and to some extent for the supply of normal AC-power from the offsite power sources, single-point vulnerabilities and human error failures in these systems have -;

~

added importance.

J NUREG-1032 - 29 -

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[NAVAILNllLITY Figure 4.5. Onsite AC System Unavailability for 18 Plants Studied In Reference 4 i

1 i

NUREG-1032 , ..

Table 4.2. Results of Onsite Power System Reliability Analysis Reported in Reference 4 Range of systen -

Diesel Generator Unavailability -

Configuration "

per Demand Dominant Tailure Causes 2-of-3 4.2 = 10 -3 _ g,g , go-2 Independent diesel railure.

- Human error CCT.

1-of-2 -

1.1 m 10 ~3 - 6.8 10 -3 Independent diesel railure.

Human error CCT. TkH outages.

3.7 = 10# 1.7 = 10-3 ." Human errei a nd' hardwar e -

2-o f-4 -

CCFf

~

1-of-3 1.8 s 10 - 7.2 = 10 " Human error, hardware.

and service water CCT.

Independent diesel failure.

DC power CCT.- --

~

2-of-5 1.4 . 10 " - 2.5 10~3 Human error, hardware ,

service water, and de power CCF.

m .

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e NUREG-1032 , ,. .._ - .. .

_~ ._ _ . _ . _ _ ~_ _ __ _ _ _ . . - _ _ _

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i Another potentially important reliability parameter involves the likelihood of restoring a failed power supply (diesel) to an operable state during a

loss of AC power transient. A histogram based on past operating experience of emergency diesel generator repair times following a failure is provided in Figure 4.6. The median repair time is approximately 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . These data represent an aggragate for all types of failure modes, and. for the mos_t

~~

l- ' part, they represent repair times during non-emergencies, p'rimarily during - -: .

plant operation but with some during plant shutdown. ---3-e-

~ ~ ' "

~ It ~1s difficult to determine whether these data over-estimate or- under- - -- -- --- - --

estimate the diesel generator repair time anticipated during an emergency.

There are reasons to believe that these data over-estimate the time to repair a failed diesel generator during a station blackout. Since the typical limiting conditions for operation (LCO) for a single diesel generator out of -

i service is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more, the urgency does not exist to restore a failed ,

I diesel in as timely a manner as would be the case during a loss of all AC -- ;

power. In addition, the LCO may not have been in force if the plant were

- '~ : shutdown ~ when a, test failure occurred, also lessening the urgency for repair. . --

And if a failure did occur, it might be seen as an opportune time to perform other routine maintenance.

Conversely, the repair time could be under-estimated by virtue of;the

~

confusion which could possibly occur during a station blackout accident. . .: _ -

h- : Un' der stress, human error is usually found to be higher than-during normal- - -

"~~onditions.

c The diesel failure problem would have to-be diagnosed, needed - - -- -

l equipment would have to be obtained, correct repair procedures would have to -

i be followed, and under time constraints and pressure all-without AC power .;

! available. Also, maintenance and operations personnel resources would be divided between activities for restoring offsite as well as emergency power supplies.

In addition to the plant-specific analyses, generic models were constructed l

. for different emergency AC power configurations. These generic models were

used to estimate system reliability as a function of the important i NUREG-1032 t

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I 3 $ 7 9 11 13 15 24 1600 ,

n si . o n. c. ...., r. ... <ni t

6

Figure 4.6 Percentage of Emergency Diesel Genera. tor Failures Repaired :.- . :

vs. Time Since Failure

'~~ ~ ~

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i NUREG-1032 .

t~. ,e y4. ,#et- -

characteristics identified in the plant-specific analyses. Nominal system dependence and common cause failure and procedural errors were assumed in the models, and sensitivity analyses were performed to determine the importance of all the factors considered. Overall, the most important factors tend to be system redundancy and emergency diesel generator reliability on demand.

NoE~sufp~risingly, it was found that common cause failure is-most important in highly redundant system configurations with highly reliable (for-independent failure causes) diesel generators. :,:, :3 - .

. - :: . c : e:, ;j-

"'i+ Based on these considerations, additional analyses of emergency- AC power -: ::- -

system reliability were performed to extend the quantitative results and .

further explore the sensitivities. Figure 4.7 shows the effect of varying emergency diesel generator reliability on emergency AC pcwer system reliability for several configurations both with and without common cause failu re. The sensitivities of system reliability estimates on variations in .

-failure to run rate are shown in Figure 4.8. Additional results-or parametric -

analyses and model development details are provided in Appendix 8.

Based on a review of past operating experience and reliability analyses, the following factors have been identified as being the largest contributors to

~

AC power system availability: -

~~ (1) The configuration of the diesel generators in terms of the number. -

~

. ,y available and the number required for shutdown cooling;3 .

' ~

2"(2) The reliability of diesel generators or other power sources used in the

~

_. . ,,-r, emergency AC power system; (3) The" dependence of the AC power system on support or auxiliary systems- - -.

used for actuation, control or cooling; (4) The vulnerability of the AC power system to common cause failure due to l various design, human error, and internal or external environmental hazards.

1 i

NUREG-1032 - -.. .

.~. _ _ _ _ __

1 With EDG Common Cause Failure .

Without EDG Comon Quse Failure ..

- Emergency AC Configuration -

' ~

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i NUREG-1032 !,

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l I In general, it has been observed that problems with onsite emergency AC power systems are very plant-specific, and improvement in system reliability would have to be developed on a plant-by-plant basis. ___5. Station Blackout Frecuency and Duration

. c;.phere. hav_e' been sev.eral incidents at nuclear power plants which could be... QT.J Z.T. Jr

classid_ __edc._.as precursors to station blackout. In fa'ct, theFelhave~ b'een-~~a'few;O=_c= -- ; 7 _. cases-iniwhich-loss-of offsite and emergency AC power supplies-occurred- .- ~

;__ __._ simultaneously. . . None of these events progressed to be a-significant safety----- ~
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li concern.- Many of these precursor incidents occurred when plants were .-. shutdown or in refueling, a condition where station blackout concerns are much reduced, and also a state where the limiting conditions for operation in terms of numbers of offsite and emergency AC power supplies available are

reduced. ^ ~ ~

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significant number of station blackout events is not surpris'ing

e lack of ~

~ ~ ~ M en.o d ~c~ensiders past frequency of loss of offsite poweEevents'-and.                                                             =--

reliability performance of emergency AC power systems. N~s~ a result it has

               . :been necessary to estimate station blackout frequency by_ combining loss of                                                            __
                                                                                                                                                        }_ - i--
                                      ~~

Tl1I-offsite powd frequency and duration correlations with t}ie emergency AC power

                     ~
               ~_re1Ta~bility.models$ Appendix B provides a description of_ithe methods-used:to-                                                                                  _

_.__._._ der 1Ve~ station 1 blackout frequency and duration estimate.s.' _ _

                                                                                                      ~ ; -- - -                              .-

3' The"results-of- sensitivity analyses performed to determiiie3hF#ffect'lif J ~ ~ ~~~ ~ ~ ~ ~

                  -design,-location, 'and emergency AC power supplies reliability _are provided in __ _. _.. .
 .__fi_ Figures .5.Lthrough 5.3. Specifically, Figure 5.1 shows!-the:effect of siteE                                                                                 ..

location and offsite power system design as represented by offsite power clusters %, 4, 5, and 7. These clusters are defined in Section 2 and Appendix A. These clusters were combined with a typical two diesel generator l 1 l l l NUREG-1032 , 4 E _

-d*

i .i i emergency AC power system with 0.975 reliability of the diesel generators. j Cluster 7 is a close representation of the average of nuclear operating experience with regard to loss of offsite power frequency and duration. . Cluster 5 represents sites with relatively high severe weather hazards and susceptibility to failure from those hazards. Cluster 4 has slightly lower

severe weather hazards than cluster 5. Cluster 2 represents the combination

    ~                                                                                                                                                        -

f l'- ~ of the more reliable offsite power design features and sites with low H- severe weather hazards or low susceptibility to severe weather; hazards.: It

q--

I ~can be ~seen that the estimated frequency of longer duration station blackouts

- 3:f f~~ is dependent on the likelihood of the more damaging and extensive losses of ;-

l offsite power for which severe weather hazards have been identified as a f i principal contributor. (Note: Seismically induced loss of offsite power has not been included but could be accounted for through a hazard evaluation and l ~ ! fragility analysis. This consideration is discussed in Section 9.) ,

        ~ Figure 5.2 shows the effect of variations in emergency- diesel generator                                                           -                            I j

i reliability for the typical offsite system (cluster 7) and emergency AC power

      ~ system (1/2 configuration). The largest change in frequency per percentile

! change in diesel generator reliability is obtained when reliability levels are lowest (0.9). This is somewhat of an artifact of the model in which common cause failure rates are kept constant. If there were no common cause -

        ~ failure contributions, or if comon cause failures were correlated with the                                                                           --

n ,- ' independent failure rate of diesel generators (and it may be), the frequency - -

e-b reduction could be proportional to the square of the percentile-change in - --

j diesel reliability for the configuration analyzed. - -- ---- - - i The last sensitivity, provided in Figure 5.3, shows the effect of emergency . AC power configuration and success criteria on station blackout frequency using a diesel generator reliability of 0.975 and a generic common cause i failure rate. Again the effect of common cause failures on system reliability j is to reduce the difference between the three configurations which would be expected from simple redundancy consideraticn!.. I NUREG-1032 l

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                                                                                                                                                                     -       39 -

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                                                                                                                                                  -           40 -

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                                                                                                                                                                              .           .t 1/3                                                              .

1 4 8 12 16 l O r St. tt .. 9/,<h.1 Duration (Hours) j 1 ! Sensitivity of Estimated Station Blackout Ouration } Figure 5.3. ! Frequency for Several Emergency AC Power Configurations NUREG-1032 i I l

Based on the station blackout analysis results, it can be seen that there is a potential for wide variation in frequency and duration depending on location, design, and reliability performance. Additional results are . provided in Appendix B.

6. Ability to Cope With a Station Blackout .

The occurrence of a station blackout is of serious concern because it has a

~= large effect on the availability of systems for removing decay-heat. -In both- -} PWRs and BWRs, a substantial number of systems normally used to cool dcwn the reactor are lost when AC power is not available. A loss of offsite power will usually result in the unavailability of the power conversion system and, in particular, inability to operate the main feedwater system. Power to reactor coolant system recirculation pumps will also be lost, requiring a - natural cir,culation cooldown to shutdown conditions. When compounded by a ,

  ' loss of the emergency AC power supplies, reactor core cooling and decay heat removal must be accomplished by a limited set of systems which are either steam driven, passive, or have other dedicated (or alternate) sources of power. Unless special provisions are made, the plant will have to be maintained in a " hot" mode, until AC power is restored. This could be hot shutdown or possibly hot standby. Table 6.1 provides a list of functions and systems for PWRs and BWRs which would be lost and-which would-remain available during a station blackout accident. Using the AC-            -

independent systems identified, decay _ heat removal can-be successful for a limited time depending on functional capabilities, capacities, and procedural - adequacy. . Decay heat removal for PWRs can be accomplished by.use of a steam-driven or dedicated diesel-driven train of the auxiliary feedwater system (AFWS). Decay heat would be rejected to the environment by the atmospheric dump valves or if necessary, by the steam generator relief valves. Since residual heat removal systems, reactor coolant make up, and reactivity control through baration would be inoperable, the plant must be maintained in a hot NUREG-1032 , _ _ _ __ _ ___

Table 6.1. Station Blackout Effects on Plant Decay Heat Removal Functions Functions (Systems) Functions (Systers) Remaining Lost PWRs Shutdown Heat Removal Shutdown Heat Removal (Motor "~ ~ - (Steam-Driven AFWS, ADVs) Driven AFWS). _. .x- . . - . . Long Term Heat Removal (RHR) . Instrumentation and Control (DC Power / Converted AC Reactivity Control (CVCS) Power, Compressed Air Reservoirs) RCS Makeup (HPIS) Pressure and Temperature .- Control (Pressurizer heaters /

                                              .            spray and PORV)

Support Systems (Service / Component Cooling Water Systems, HVAC, Station Air Compressors) . eg : _

      ~BWRs      Shutdown Heat Removal                   Long Term. Heat Removal (RHR)

(2/3) (Isolation Condenser, Fire ._.. . WaterSystem) RCS Makeup (LPCS, FWCI) Instrumentation and Control Support Systems (DC Power / Converted AC (Service /ComponentCooling Power, Compressed Air Water Systems, HVAC, Station Reservoirs) Air Compressors) NUREG-1032 . _

Table 6.1. Station Blackout Effects on Plant Decay Heat

Removal Functions (continued)

Functions (Systems) Functions (Systems) Remaining Lost ._ e'~ -i BWRs Shutdown Heat Removal Long Term Heat . :. - -: - - (3-6) & RCS Makeup Remoyal -:

                  ~:(HPCI or HPCS/RCIC)                           (Shutdown Cooling System, Low Pressure Coolant                ,

Instrumentation and Control Recirculation System, (DC Power / Converted AC Suppression Pool Cooling Power, Compressed Air System) Reserviors) , Support Systems (Service / Component Cooling Water Systems, HVAC, Station Air Compressors)

                                                                                                    -   i i

I l ! NUREG-1032 .

condition. The plant operating state (primary coolant pressure and _ temperature) would be maintained by manual operation of the AFWS and I atmospheric steam dump valves. With primary coolant pumps unavailable, - . l reactor core cooling would be achieved through natural circulation. If the AFWS can remain operable, and if primary coolant inventory can be - maintained at a level adequate to maintain the core cooling / heat transport P~ ~ loop to the steam generators, a PWR should be able to stay in this-mode of- .- decay heat removal for a substantial period of time. The amount of. time that

        ' decay heat removal can be maintained in a PWR is generally limited by primary pressure boundary leakage and the capacity of certain support or auxiliary systems. The sources of potential leakage include reactor coolant pump seals, unisolated letdown lines, and a stuck open pilot operated relief valve (PORV). With provisions for manual isolation of letdown lines and reduced frequency of PORY demands, the reactor coolant pump seal leakage rate is          ,
        - considered to be a potentially limiting factor for some designs. If the leakage rate is low, on the order of several gallons per minute, this concern is negligible. However, if seal leakage is on the order of 100 gpm or more, reactor coolant system inventory depletion will be a factor limiting decay heat removal for an extended period of time.

Natural circulation cooldown in PWRs has been successfully demonstrated by actual operating experience. The process becomes more difficult with AC power unavailable because reactor coolant makeup to accommodate system shrinkage and pressurizer heaters or sprays to help control coolant conditions are inoperable. Nevertheless, analytical evaluations (Ref. 7) and experimental observations (Ref. 8) show that decay heat removal can be achieved with the system operational limitations associated with a station blackout. In fact, core cooling is expected to preclude core melting even with significant voiding in the primary coolant system if the steam generator is maintained as a heat sink. For station blackout purposes, BWRs have been divided into two functionally different classes: (1) those that use an isolation condenser cooling system NUREG-1032 for decay heat removal and do not have an AC-power-independent makeup capability (BWR-2 and -3 designs), and (2) those with a reactor core isolation cooling (RCIC) systen,and either a steam-turbine-driven high pressure coolant injection (HPCI) system or high pressure core spray (HPCS) system with a dedicated diesel, any of which are adequate to remove decay

           ' heat from the core and control water inventory conditions in the reactor                  ,-
           - vessel (BWR-4, -5, and -6 designs). Since BWRs are designed as natural
 'M        M circulation reactors, at least at reduced power levels, .the- loss of reactor-  -
                                                                                                    ~-

7 - ' coolant ~ recirculation poses no special consideration.- Moreover, reactivity- -. control during cooldown is adequately maintained by control rod insertion, an action which would occur automatically on loss of all AC power. The isolation condenser BWR has functional characteristics scmewhat like that of a PWR during a station blackout in that normal make up to the reactor coolant system is lost along with the residual heat removal (RHR) system. ,

             'The' isolation condenser is essentially a passive system which is actuated by opening a condensate return valve and transfers decay heat by natural
       '~

circulation. The shell side of the condensor is supplied with water from a diesel-driven pump. However, replenishment of the existing reservoir of water in the isolation condenser is not equired until one or two hours after actuation. It may also be possible to remove decay heat from this class of

             ' BWRs by depressurizing the primary system and using a special connection for      -    -
          - t a fire water pump to provide reactor coolant make up. This alternative would     .

require a much larger operator involvement. - A large source of uncontrolled primary coolant leakage will limit the time-

for which the isolation condenser cooling system can be effective. If no , source of make up is provided, eventually enough inventory will be lost to uncover the core. A stuck open relief valve or reactor coolant recirculation pump seal leakage are potential sources of such leakage. The necessity to maintain the operability of auxiliary and support systems such as DC power and compressed air is less than for the PWR when isolation condenser cooling has been established. However, these systems would eventually be required to I recover from the transient. l NUREG-1032 The BWRs with RCIC and HPCI or HPCS can establish decay heat removal by discharging steam to the suppression pool through relief valves and by making up lost coolant to the reactor vessel. In these BWR designs, d'ecay heat is not removed to the environment, but is stored in the suppression pool. For this type of BWR design, long term heat removal in the form of

                       - suppression pool cooling or RHR using low pressure coolant. injection and                        -

recirculation heat transport loops is lost during a station blackout. The .

                        . duration that the plant can be maintained in a safe condition. prior to AC      .
                       - power recovery is determined, in part, by the maximum. suppression pool        .

temperature for which successful operation of decay heat removal systems can be assured both during a station blackout transient and upon AC power recovery. At high suppression pool temperatures (around 200 F), unstable condensation loads may cause loss of containment suppression pool integrity. Another suppression pool temperature limitation to be considered is the qualification temperature on the RCIC or HPCI pumps to be used during . recirculation. Also, net positive suction head (NPSH) requirements for pumps in systems required to effectuate recovery once AC power is restored may limit suppression pool temperatures. . In general, all LWR designs have the ability to remove decay heat for some period of time. That time period is dependent on support system capabilities and capacities such as the quantity and availability of water. sources . . . required for decay heat rejection, the capacity of DC power supplies and compressed air reservoirs, and the potential degradation of components due to > environmental conditions arising in the absence of HVAC systems. System capabilities and capacities are normally set to provide safety function operability for the spectrum of design basis accidents and anticipated operational transients, of which station blackout is not one. Perhaps the most important support system for both PWRs and BWRs is the DC power supply. During a station blackout, unless special emergency systems are provided, battery charging capability is lost. Therefore, the capability of the DC system to provide power needed for instrumentation and control can NUREG-IO32 , 47 _

3 be a significant time constraint on the ability of a plant to cope with a station blackout. DC power systems are generally designed for a certain capacity in the event of a design basis accident with battery charging unavailable. However, the system loads for decay heat removal during a total loss of AC power are somewhat less than the expected design basis accident loads on the DC power system. Therefore, most DC power systems in operation ,- today have additional capacity to last for a longer time during a station blackout than would be expected for a design basis accident.- - --

Another important item for decay heat removal during station blackout is the - condensate storage tank capacity. Normally, this tank contains a sufficient .- amount of water to cool the reactor until the RHR system can be placed in operation. Since the RHR system is unavailable with a loss of all AC power, the ability to cope with station blackout is a function of the condensate storage tank capacity. The ability to provide makeup to'the condensate ~ storage tank with systems and/or components independent of station AC power would extend this potentially limiting factor. Also, during a station blackout, there may be need to operate scme pneumatic valves, such as the steam relief valve. Since AC power is not available, the station air compressors will be lost. For this reason, local air reservoirs are normally provided which will pennit the valves to be operated for a-C limited number of cycles. After exhaustion of the air supply, operation of . these valves may have to be manually performed by the operations staff. .

l I During a station blackout, normal heating, ventilation and air conditioning in the plant would be unavailable. Equipment needed to operate during a .

station blackout and equipment required for recovery from a station blackout l would have to maintain operability in the environmental conditions (e.g.,

! 9emperature, pressure, humidity) that could occur as a result of the event. Otherwise, failures of necessary equipment could lead to loss of core cooling , and decay heat removal during the event or failure to recover from the event upon recovery of AC power. The instrumentation and control elements of components required during station blackout are the most likely to be impacted by adverse environments. However, only a limited set of equipment NUREG-1032 -

located in the control room would need to be operable, thus reducing equipment-generated heat loads in that location. The same would be true for equipment . located in auxiliary buildings and inside containment, although sensible heat from preexisting sources could be considerable. For control rooms and auxiliary buildings, opening doors should provide adequate relief to

                                                                                                                                   ~
       '~' maintain equipment in an acceptable operating environment. Most equipment in-                                                   .

containment is designed to function in the more limiting environment associated - with a design basis loss of coolant accident, and therefore,-could be expected- - :

to function during a station blackout transient. A summary of design-related factors which have been identified as potentially limiting the capability of LWRs in coping with a station blackout is provided

in Table 6.2.

Actions necessary to operate systems that are needed to establish and

              ~~ maintain decay heat removal and fully recover frcm a station blackout would                                                    -

not be routine. The operator wnuld be presented with significantly less information and operational flexibility than is normally available during most other transients requiring reactor cooldown. On the other hand, the loss of all AC power is an easily diagnosed occurrence, although not always easily corrected. -

Operational staff activities would have to be directed at both. reactor decay -- - This would involve

heat removal requirements and restoration of AC power. : manual operations within the control room to control the rate of core decay heat removal and performing special activities outside of the control room. The latter activities would include repairing of failed components, isolating sources of reactor coolant leakage, conserving DC power through load 4l stripping, making available alternate make up water supplies, hooking up compressed air bottles, and possibly local manual operation of some components. The success of these activities would require preplanning,

training and procedures. In addition, adequate lighting and communication must be available. Where local access is necessary, security and working l environment (pressure, temperature, humicity, and radiation) could be limiting factors.

NUREG-1032 ._ _ - - - - - - _ - . -_. _ _ . - _ . - - _ - _ - _ - _ . - , - - --_. - _ .

Table 6.2. Possible Factors Limiting the Ability to Cope with a Station Blackout PWR BWR 2/3 BWR 4/5/6 RCS pump seal leakage X X .- RCS letdown / makeup & water X X; e:. ; . . chemistry control lines -.- Stuck open relief valve X X DC battery capacity (18C) X X X Compressed air (valve control) X X X j DHR water supply X X X

 >   (condensate, firewater)                                                                                          .-

Aperating environment . (temperature) - control room (I&C) X X X containment X (suppression pool, l wetwell, drywell) auxiliary building X X (AFWS room) (HPCI/RCIC room) .- T i i i NUREG-1032 l

In PWRs, operators must control the rate of heat removal from the steam generators by the auxiliary feedwater system in order to maintain the proper pressure and temperature balance within the primary coolant system to allow for adequate natural circulation and to maintain water level in the pressurizer, Although there is analytical and experimental evidence that suggests that natural circulation and adequate decay heat removal can be maintained when . pressurizer level is lost, and, in fact, when a two-phase flow mixture exists in the reactor coolant system up to the point the reactor core is-uncovered, these conditions would complicate the recovery process adding-to the difficulty of operator recovery actions. In the BWR, the isolation condenser appears to need less attention by the operator. However, one would have to assure that automatic depressurization does not occur and that the makeup system to the isolation condenser is operating properly within approximately 2 hours of the loss of AC power. In BWRs with HPCI or HPCS and RCIC, the operator must control both pressure and the level of reactor coolant in the vessel. This . requires actuation of both makeup and relief systems. In all LWRs, operators would have to be prepared to deal with the effects of losing and restoring AC power on plant control and safety system set points to limit additional transient complications and assure operability of AC-powered cooling systems.

7. Accident Seouence Analyses Accident sequence analyses have been performed to determine the accident progression character'stics (Ref. 7, 9-11) and likelihood (Ref. 5) of a station blackout. Using fault trees and event trees, accident sequence functional and system failure characteristics were identified. Reactor coolant system transient response analyses were used to detennine the capability of coping Dith station blackout and to estimate the time available for AC pcwer recovery prior to core damage and core melt for the potentially important functional failures during a station blackout.

In light of the decay heat removal system capability requirements, and considering the reliability and failure modes and causes of failure of these systems, three phases of a station blackout transient were identified. The first includes the need for promptly actuating decay heat removal systems and NUREG-1032 2he potential for a station blackout induced loss of coolant accident (LOCA), either'of which can result in a loss of core cooling within li to 2 hours. The second phase lasts up to approximately 8 to 12 hours and includes operational , limitations in the capability of continued decay heat removal considering limited capacities (e.g., DC power, condensate storage) or interactive failure (e.g'., high temperature effects due to loss of HVAC), and the potential for reactor coolant loss (e.g., pump seal leakage). System running reliability is ~ ' of~ lesser importance during this time period compared with the successful - - -. - ~ initial actuation of systems to remove decay heat. The third time-interval -

                                                                                             -~7:

thich extends beyond 8 to 12 hours, involves the need to eventually recover AC power and establish a stable, controllable mode of decay heat removal. Consideration of the systems and functions available for the different PWR and BWR designs, as discussed in the previous section, resulted in the development of three event trees for the identification of station blackout . accident sequence scenarios. These are provided in Figures 7.1 for PWRs, 7.2 for BWRs which use an isolation condenser, and 7.3 for BWRs which have AC-independent make up systems (RCIC, HPCS, HPCI). The event trees are characterized not only by the systemic and functional considerations important to station blackout accident sequences, but also by considering the phases of the transient which would affect the plant response and system operability associated with the duration of various station blackout sequences. The event trees begin with the loss of all. AC power as the initiating event and proceed thru decay heat removal, reactor coolant - inventory (integrity), and restoration of AC power to operate the normal . decay heat removal and makeup systems. The accident sequence logic is similar for PWRs and isolation-condenser BWRs which do not have the , capability of making up lost reactor coolant during a station blackout. These plants are susceptible to degraded core cooling for relatively small leakages of reactor coolant. BWRs with reactor coolant makeup available during a station blackout have a somewhat different accident sequence logic. Most losses of reactor coolant caused by station blackout can be accommodated by the available reactor coolant injection systems. Reactor coolant leakage equivalent to a stuck open relief valve can be accomodated by the RCIC systems. The HPCI or HPCS system can provide adequate make up to cope with NUREG-1032

-2 hre. 2-12 hrs. >12->24 hrs.

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larger leaks. All of the LWRs encompassed by the accident logfc models are subject to operational limitations for the longer duration blackouts as described previously in Section 6.

The event trees end with a sequence outcome state designated as "OK," meaning stable, long-term core cooling is achieved or achievable, or "CD." meaning an

                                                                                                         .f-inadequate core cooling state is reached and some reactor core damage can be
  ~;     ~

expected. For the latter case, core damage can be expected to proceed to core :- --:

' melt if effective and timely restoration measures for AC power and core cooling ~.- --

are not taken or available. The difference between an accident sequence ending

!           in core damage versus one which leads to core melt is determined by evaluating the likelihood of restoring core cooling and its effectiveness from the time of onset of core damage to the time at which irrevocable core melting has begun.

- This latter time in the accident sequence progression is not well known due to sigr.ificant uncertainties in the modeling of core melt phenomena. It has been estimated that the time difference between the onset of core damage to the time - that a core melt would penetrate the reactor vessel is on the order of 1 to 3 hours (Ref. 6, 9). Considering the uncertainty in modeling this accident - - process, including the ability to terminate a core melt in progress, it has been assumed that core melt would be the likely final outcome in accident sequences which progress to core damage. Detailed plant transient response analyses were performed to cover the spectrum - of sequences identified in the event trees (Ref. 4-6). The purpose of this work was'to better understand accident progression characteristics related to- -- the timing of events and physical parameter values during-the transient, and to determine success states for systems, trains, components -and operator actions , during station blackout sequences. The sequences were divided into three groups, namely: (1) failure of AC-independent decay heat removal with reactor coolant ! leakage less than technical specification upper limits; 4

(2) failure of reactor coolant system integrity (liquid or steam leaks) with AC-independent decay heat removal systems operable; and NUREG-1032 .

(3) failure of AC-independent decay heat removal systems with loss of reactor coolant system integrity. Variations in system failure and actuation time, reactor coolant leak rate, and operator actions were analyzed to determine the potential for sequence outcomes involving adequate or inadequate core cooling as well as. the .. time constraints on recovery actions to avoid core damage. .

Results provided in Table 7.1 show the estimated time of. core uncovery for m....,..

station blackout sequences with AC-independent decay heat removal systems . . unavailable. Plants with B&W type NSSS which have a small steam generator secondary water inventory and, thus, the smallest heat capacity would require the most prompt recovery to avoid core damage for this particular sequence. Core uncovery was estimated to occur within one hour. Plants with Westinghouse or Combustion Engineering NSSS designs show a somewhat longer time to core . r- uncovery, about two hours. A BWR-4 type of plant shows core uncovery in about .

!                two hours also. Figure 7.4 shows how the core uncovery time is extended for.

sequences in which decay heat removal is initially successful but fails at a. .: later time during the accident. An estimate of the time to core uncovering with a stuck open relief valve and other types or reactor coolant leakage are

,                also provided in Table 7.1. For BWRs with RCIC available (or HPCI, HPCS),

adequate reactor coolant make up is provided to maintain core cooling with a ! stuck open relief valve. It was also found that the core uncovery. time of PWRs , ._._ t:as not significantly shortened when a stuck open relief valve compounded the . 1 ,' - loss of the steam turbine-driven train of the AFWS. This is because. loss of. decay heat removal usually results in primary system pressure relief to remove decay heat almost equivalent to the energy loss of a stuck open relief valve I with AC-independent decay heat removal available. EWRs without RCIC or with AC-independent decay heat removal systems unavailable have a somewhat shortened 1 time to core uncovering with a stuck open relief valve. ; Complete accident progressicn analyses have been performed for several key , ! station blackout sequences starting with the loss of offsite power through to f i I f NUREG-1032 l l l r

T 5 Table 7.1 Estimated Time to Uncover Core for Station Blackout _ Sequences with Initial Failure of AC-Independent 4 Decay Heat Removal Systems a'nd/or Reactor Coolant . Leaks Core Uncovery Time (seconds) - PWRs i i Sequence B&W CE W AFW fails 2715 6200 5800 Stuck Open PORY 3190 .

                                                                     -             5040 NactorCoolantPump               21070                           -            27950
;    Seal Leak Rate of                                                         ,,      ,

100 gpm total AFW fails and Stuck 2480 - 4800 Open PORY - i BWRs GE + HPCI/RCIC fail 2300 I 2 HPCI/RCIC fail and Stuck 1680 l Open SRV i NUREG-1032 < .

e Loss of offsite power, failure of all diesel generators, technical _ specification leakage, turbine driven auxiliary feedwater initially operates then falls at a later time.

                          !              .           I                                   I                                    I                        .

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O 5 10 15 20 25 T!se of felture (HR) i Figure 7.4. Time to Core Uncovery as a Function of Time l at Which Turbine-Driven Auxiliary Feedwater Train Fafis I l 1. r NUREG-1032 .

core melt and containment failure. A time line presentation of a PWR sequence in which AFWS operation is initially successful but fails several hours into the transient is provided in Figure 7.5. A station blackout occurs at zero hours (t0 ). After the initial fluctuations in reactor coolant system pressure core outlet temperature, pressurizer level, core flow, and steam generator, a relatively stable period of decay heat removal with primary coolant natural ,. circulation follows. When AFW make up to the steam generator becomes unavailable in about six hours (t y), steam generator level begins to drop '

   ' causing decreased heat transport from the primary coolant system. . As the steam
 ~                                                                                        - - - -

generator dries out and heat transfer to the secondary system ceases, reactor coolant pressure and core outlet temperature rise. Reactor coolant temperature increase combined with some voiding causes the pressurizer. level to rise, and relief to containment occurs. Continued voiding in the primary system affects core natural circulation flow, but core cooling is adequate to prevent melting until it is uncovered (t )2 at about.nine hours.. At this point the pressurizer _ level has dropped since most of the primary system is voided. Within about two more hours (t ) the core has melted and penetrated the reactor vessel causing a 3 containment pressure and temperature spike due to the-rapid influx of steam and noncondensables frcm the melt. If containment survives that spike, the continued release of decay heat and the generation of combustible and non-combustible gas will continue to load the containment. Containment failure by overpressure is shown here at about 19 hours into the-accident. - A BWR station blackout accident sequence progression is shown in Figure 7.6. In this BWR with Mark I containment, a station blackout occurs at - time zero (t0 ). The reactor coolant system pressure ano level are maintained within limits by RCIC and/or HPCI and relief valve actuations thich also transfers decay heat to the suppression pool. Both the suppression pool and drywell temperature begin to rise slowly, the latter being more affected by natural convection heat transport from the hot metal (vessel and piping) of the primary system. Af ter one hour, when AC power restoration is not expected, the operator begins a NUREG-1032 i

D;I;ysd failuro of AFWS (or DC power depiction) 1 Rzctor Coolant System Pressure l L Pressurizer Level k ( . Core Flow e

i. -.-

Core Outlet Temperature Stenm Generator La' vel [ F^ - Conta.inment . Pressure 7 Containment - Temperature 7

                                                           ,                 i                             !

8 .12 16 20 Time (hrs) 0 4 t4 - to tt t2 t3

                                                                                                        ~
                                                                                                                    ; -  ~
   .                   Time                   Sequence Event Loss of all AC power                            .

to AFWS fails (or DC power depleted) ti Core uncovery begins

                                                                                ~                                               ---

t2

                                    -         Reactor vessel penetration t3 t4                   Containment failure Figure 7.5. PWP. Station Blackout Accident Sequence NUREG-1032                              i

RCIC/HPCI cvcIlablo, controlicd depressurization Reactor Vessel Pressure Reactor Vessel Level Core Temperature T - . - .. . Suppression Pool Temperature Drywell Temperature . f _ . _ - . . Drywell Pressure ' t , 4 8 12 16 Time (hrs) 0 to t3 t2 t3 t4 l l l Time Sequence Event . Loss of all AC power to DC power (batteries) depleted - ti

Core uncovery begins ,

t2 Reactor vessel penetration t3 t4 Containment failure Figure 7.6. BWR Station Blackout Accident Sequence f NUREG-1032 '

controlled depressurization of the primary system to about 100 psi. This also causes a reduction in reactor coolant temperature from about 550 F to 350U F which will reduce the heat load to the drywell as primary system metal components are.also cooled. The suppression pool temperature increase is only slightly faster than it would have been without de~ pre ~ssdrization. Drywell pressure is also slowly increasing. At about 6 . . _ - .m _ hours, DC power supplies are depleted and HPCI and RCIC are no longer . operable. Primary coolant heatup follows with increases:in ~ pressure and . level-until-the safety relief valve set point is reached. Continued core "heatup causes continued release of steam, eventually depleting primary coolant inventory to the point that the level falls to uncover the core about two hours after loss of makeup (t2 ). Core temperature begins to rise rapidly resulting in core melt and vessel penetration within ano+Mr two or three hours (t3 ). During the core melt phase, containment pressure and temperature rise considerably, such that nearly. coincident with vessel , penetration, containment failure occurs, either by loss of electrical - penetration integrity (shcwn at t )4 or by containment over pressure shortly thereafter, around 11 hours into the acciderft. Accident sequence likelihcod estimates were made to identify the potentially dominant station blackout accident sequence contributors (Reference 5). Table 7.2 provides a summary of the results for the typical PWR and BWR, These results have been modified to account for better estimates of loss of offsite power frequency and duration derived since the completion of that ;. work (seeAppendixA). In addition to the dominant accident sequences and their likelihoods, the major factors affecting the accident sequence frequency .

are also identified. For PWRs, an important contributor to core damage likelihood estimate is the ability to restore AC power prior to the loss of DC power needed to run the auxiliary feedwater system or the depletion of condensate storage tank supplies. Another important contributor involves the integrity of the reactor coolant system considering the reactor coolant pump seal leakage potential following a station blackout accident. If reactor coolant pump seals leak without the ability to make up to the reactor coolant system, the ccre will be uncovered. Very large reactor coolant seal leakage (greater than 100 gpm per pump) could result in core uncovery within a few NUREG-1032 6

f4;' i f 1 1 h Table 7.2. Sumary of Potentially Dominant Core Damage Accident Sequences E. ii. . . . -i i . y Generic AC Recovery Time Typical Core m Plant Sequence DIR System / Component Contributors to Avoid Core Damage Damage Frequency PWR THijj 8 5' team,',' driven AFWS. Mavailable 'l l i! ' huo' rs 5 x 10-6 (all) THL22 8 DC popr or condensate exhausted 4-16 hours 1 x 10-5 THQ 822 RCS pump seal leak 4-16 hours 1 x 10-5 BWR Thug 8j Isolation condenser unavailable 1-2 hours 2 x 10-6 w/ Isolation Condenser - ha THQy8; Stuck open relief valve 1-2 hours 3 x 10-6 RCS pump seal leak 2 x 10 -5 THQ 822 4-16 hours q .._.................__ __....... ......... .....__....

l ;: '.* BWR THUj8y HPCI/RCIC Unavailable 1-2 hours 2 x 10

                                                                                                                                                                     '.c w/HPCI-RCIC
                                                           .                                          4 THU22       8         DC power or condensate exhausted,                   4-16 hours                       2 x 10'5 component operability limits exceeded (HPCl/RCIC)                                     e,,             ,

i- ,,,i.i..

t BWR THb{B y HPCS/RCIC unavailable 1-2 hours 5 x 10~7 w/HPCS- i - i RCIC

,u

u. -g THU22 8 HPCS unavailable, DC power or 4-16 hours 1 x 10 condensate exhausted, component operability limits exceeded
i (RCIC)' ' t l

e p , e e e a Y lg , .I?fa t h I {,

           ..    - - -      -                    --                     - ~ - . . . _ -             --          - -

j . . . 1 j . . hours. Smaller leakage rates (a few gpm per pump) do not represent a limiting f factor. ' Adequate coolant inventory wguld be available to allow continued core !- cooling for a day or more without the need for make up if other limitations ! (e.g.,DCpower)didnotexist. In the analyses performed for this program, j the reactor core was found to be uncovered in approximately 8 hours using the

reactor coolant seal leakage information currently available (i.e., a leak rate .

':            of about 10 to 20 gpa per pump)?                                          .,    _                             :l i-                      -      a                                                            - :  -        -   ;            i b:                             a                                                      a  .,               .

[, For BWRs with isolation condensers, a similar dominant . failure ~ mode ex1sts.- . l The failure of the DC power system is less important because the isolation j condenser system operates passively once initially activatsd, with little ! j operator action necessary thereafter. However, reactor coolant pump seal

failure could cause depletion of reactor coolant inventory, and since the isolation condenser BWR typically does not have an AC-power-independent .

[ makeup system, the reactor core could be uncovered. This sequence was , j'i estimated to result in core damage in about an 8 to 12, hour time frame. : l _ ~BWRs.with HPCI and RCIC are capable of coping with reactor coolant , system leaks equivalent to a stuck open relief valve. '.However, they are - i

-}

j subject to DC power depletion effects and other interactive failures - associated with the, lack of the ventilation system to maintain HPCI and

i RCIC. room temperature, and suppression pool heat up due to loss of. 1

       ;~ cooling can resultiin a loss of core cooling in the latter time frame of i                       ;             J Q,              about_8-12 hour {s.7 For this type of plant, unattenuated suppression pool              _   _

j l temperature increases during a station blackout transient can be a

problem. This is due to the potential for unstable condensation j phenomena which could cause containment structural failure with the- - f.

poteritial loss of reactor coolant to suppression pool recirculation r [ { capability. Perhaps more important is the effect that high suppression j pool temperature would have on HPCI pumps during recirculation. These ! pumps are not usually qualified for operation with fluid temperatures in [ excess of 160*F. In addition, net positive suction head requirements may not be satisfied if suppression pool temperatures exceed 200*F.

For BWRs with HPCS, which has its own AC and DC power systems, the DC power l supply depletion effects, as well as reactor coolant leakage failures, i

! NUREG-1032 !

i . . i l t i l are minimal contributors to sequence core melt probability. However, ) suppression pool temperature limitations may cause some equipment . l operability problems during longer duration station blackouts. , In all of the accident sequences, evaluated for this program, the early [ failure of decay heat removal due to initial unreliability of these , l systems was found to be a relatively small, but not insignificant, l j~: ' contributor to core melt frequency. This is not surprising, since . i +C following the TMI accident, most nuclear power plants-have been-required - .-- - --

-e

, to have at least one AC-power-independent decay heat removal train .- . - I available. On the other hand, a very limited amount of work has been < I done at nuclear power plants to look at the capability and reliability of systems in conditions of sustained loss of AC power. Therefore, it i is not surprising that most of the dominant failure modes that have been identified have been associated with the inability to operate decay heat

removal systems because of support system failures or capacity limits on [' support and auxiliary systems needed to maintain decay heat removal l during the station blackout. : i l With the consideration of containment failure, station blackout accidents can )^ represent an important contributor to reactor risk. In general, active l' 1

containment systems are unavailable during a station blackout accident. - - i These systems are usually required for pressure suppression;through

+ steam condensation to maintain the containment pressure below the - - - . --

j' :: appropriate limits and for radioactivity removal from the-containment- >- atmosphere following an accident. The time to containment failure ; l ' ~ following the onset of core damage and the containment failure mode ,j i represent important factors which influence fission product release, and j ultimately public risk. l i

i Containment failure insights derived from analyses performed for this program j and a survey of analyses performed for other programs are summarized in Tabl'e 7.3. Identified are the different types of containment, the estimated l

j time of containment failure following the onset of core melt, and the , NUREG-1032 l L ..

Table 7.3. Containment Failure Insights Approximate Time to Containment Failure Following Onset of Most Probable Containment Cantainment Type Core Damage Failure Modes . Ic3 Condenser 1 hr. Hydrogen burn, steam spike. .. ,

   ,g            3                               2 hrs.                     Overpressure .           -

At or following AC Hydrogen burn ' ' recovery

g .,, , y ,. , , 5 ee R efer en< < ss Substmospheric 2 hrs. Hydrogen burn, steam or Small Dry spike 6-12 hrs. . Overpressure Following AC HyNrogen burn recovery
Large Dry 10 hrs. Overpressure Following AC Hydrogen burn recovery
3e urs, see Refer en e 13 Mark I, Mark II 2-4 hrs. Electrical penetration failure .. . -

4-8 hrs. Overpressure s1 krs, s'e e Referen s

I3 Mark _ III 10-15 hrs. Overpressure ,, ,

1 hr. following Hydrogen burn AC recovery " 49 hes, Cea geferegee t *3

        'Ce ends en accident . anagesent stratag for hydrogen c:ntr:1.

NUREG-1032 i l _- .-. ,, . . _ . -. - :"#~_ .-- _" ~** K -- - ~l ~' --

                                                                                                       ~.---- . - - -   -- --

type of containment failure mode. The most recent estimates of containment performance derived from ongoing severe accident research by both NRC (Ref.12) ! and the Industry Degraded Core Rulemaking Program (IDCOR) (Reference 13) may be , cause for revision of containment performance insights derived just a few years previously. For the large dry type of PWR containment, long-term overpressure is the most likely failure mode. Yet some evidence exists that some very , strong large dry containments may not fail at all due to overpressure for 2' station blackout accidents, since they have capability of withstanding the J

~ ' " o~ erp~ressure     v            transient. The smaller PWR containments, like the subatmospherie           2     ." .

[ or the ice condenser designs with lower design pressure and smaller volume, are less capable of handling the pressure transient and potential hydrogen burn i associated with a station blackout core melt accident. Containment failure in about 1 or 2 hours has been estimated to occur by several possible failure i modes including hydrogen burn, steam pressure spike, or containment overpressure due to non-condensables and non-condensed steam. Again, the . recent IDCOR results show containment failure times of more than one day. --

                    - The BWR Mark I and II containments offer some pressure suppression capability during a station blackout accident, but following a core melt their failure may occur by one of two modes. Either mechanical or electrical fixtures in the penetrations may fail because they are not designed for the pressure and temperature which will follow, or ultimately, in about 5 to                             -

8 hours, overpressure of the containment will occur. IDCOR estimates are on - - 4 the order of 18 hours for a Mark I containment. Since these containments are --- l generally inerted, hydrogen burn is not considered a likely failure mode. For. - Mark III containments, which are icw pressure but large volume containments, failure in 10 to 15 hours has been estimated, principally by overpressure. . I The IDCOR estimate is 47 hours for this type of containment. , i , One item of interest that should be noted is that for both the ice condenser containment and the Mark III containment, where the installation of hydrogen ignitors has been required to meet hydrogen rule requirements and the post l Construction Permit Manufacturing Licensee (CPML) rule, the potential exists

i. for turning on an inactive ignitor following the restoration of AC power at a time when the hydrogen concentration is essentially at an explosive mixture.

NUREG-1032 _ _ _ _ _ _ - - _ - . _ .__

i j This accident management consideration can be influenced by accident procedures [ and information available to the operators to control the hydrogen burning with j the restoration of AC power and ignitor systems. L i In Table 7.4 the fission product release categories have been correlated with ] the different containment types and failure modes identified in Table 7.3. -j ; 4 The doses estimated to result from station blackout accidents for the various different containment designs are provided, including recent IDCOR estimates. l Although substantial uncertainties exist regarding fission product transport in containment during a core melt accident, it can be seen that station blackout j accidents can potentially result in substantial fission product releases based [ on an understanding of the fission product transport process as known today. Again, the reader is cautioned that ongoing research could cause substantial i revision of these fission product release fractions shown in Table 7.4'. , i 8. Evaluation of Dominant Station Blackout Accident Characteristics , i I Baced on the previous work presented on dominant station blackout accident l secuences, the important factors which affect the probability of station [ blackout accidents have been identified. The principal parts of the station j blackout sequence include: the likelihood or frequency of loss of offsite j power; the probability that the emergency or onsite AC power supplies will ( , be unavailable; the capability and reliability of decay heat removal systems f which must function during a loss of AC power; and the likelihood of ! restoring a source of offsite power prior to the onset of core damage due f to loss of core cooling and failure of systems which cannot operate [ without AC power available. The reactor type by itself has not been found to be a dominant factor in determining likelihood of core damage } due to station blackout. This is because capabilities of auxiliary and support systems needed for decay heat removal during station blackout can vary considerably and still meet current safety requirements. The i reliability of the AC pcwer system (offsite and onsite) and the ! I i NUREG-1032 i

                  ._. _ --   _ _ . _. _    _   . . _ , _ . - - _ _ _ . _-     _ ._, ._ _   E.

Table 7.4. Containment Fission Product Release Categories and Failure Mode Probabilities for Station Blackout Sequences Release Containment Type - Category Failure Probability and Mode BWR Mark Ice Sub-Atmos- Large Dry- Large Ory-1, 2/3 Condenser pheric Wet Cavity Dry Cavity ,- 1?- 1E-4(a) 1E-4(a) 1E-4(a) 1E-4(a) . .:1E-4(a) :. 2 - .2(y') --

                                               .1(6,)         .8(6,)                .2(6,)                 -

3 .8(y) .99(60 ) .9(6j ) .2(6j ) .2(6)) 4 -- 1E-2(s) 1E-3(s) 7.3E-3(s) 7.3E-3(s) 6 -- -- -- --

                                                                                    .6(c)

Total 5.5 x 106 / 5.3 x 10 6 5.4 x 10 6 4.9 x 10 6 2.1 x 10 6 Man-rea 4.5 x 10 Out to 50 Mi IDCOR 1.3x10f/ 7.3 x 10 5 -- -- 8.2 x 10 5 Results 2.4 x 10 Containment Failure Modes -- a - steam explosion . ; y' - overpressure, direct atmospheric release ,--- m -- r , y - overpressure, release through reactor building - 6 - overpressure, late 7 6, - overpressure due to steam spike at time of vessel melt thru 6 0 verpressure with core debris bed fragmentation l s - containment leakage c - base mat melt thru l l l 1 NUREG-1032 ~

performance of these auxiliary systems (DC power, compressed air) in addition to plant characteristics (pump seal design, natural circulation capability, suppression pool temperature effects) will be important determinants of station blackout core damage likelihood. It' was recognized that core damage frequency estimates for station blackout occident-sequences could vary considerably based on differences in these factors. Therefore, the sensitivity of core damage frequency es.timates to design variations different from the reference plant analyses performed by I Sandia National Laboratories (Ref. 5) was analyzed. The models used to perford'

the analysis are based on insights obtained from previous studies and are described in Appendix C. Station blackout sequences have been divided into two groups. The first includes sequences involving the failure of AC-independent decay heat removal or, for plants without'AC independent makeup, loss of reactor coolant integrity at the onset or soon after a . station bl.ackout. For _ _ . . I these early core cooling failure types of sequences, AC power must be restored in one or two hours to avoid core damage and ultimately core melt. The second group of sequences identified includes failures during an extended station blackout of 4 to 8 hours or more. These failures include a smaller rate of reactor coolant loss, support system capacity limitations (e.g., batteries,

 --   make- up water inventory, compressed air), or other station blackout capabil.ity__ _           _

l limitations in decay heat removal systems (e.g., natural circulation and

suppression pool temperature limitations).

                                                                     . i: ' _

Several sensitivity analyses have been performed to evaluate = variations in= - ~ { LWR plant designs for both decay heat removal capability and system _ _ l reliability including offsite power. Because ability to cope with a station blackout may vary considerably, results are provided to show the affect of limitations in maintaining decay heat removal during station

blackout durations of 2 to 16 hours. First, Figure 8.1 shows the sensitivity to offsite power system design and location as represented by different offsite power groups (clusters). The importance of higher { frequency and long duration losses of offsite power can be seen. It is NUREG-1032 l

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                                               = m_ww- ;=w.n=w.,mmem                                                                                                                                       OHR Reliability

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10'g - 8 12 16 0 4 Station Blackout Capability (Hours) l Figure 8.1. Sensitivity of Estimated Station Blackout-Core Damage Frequency to Offsite Power Cluster, AC-Independent Decay Heat l l Removal Reliability, and Station Blackout Coping Capability I NUREG-1032

                                                                                                                                        .        72

[ l _ _ ,

                                                                                                                                                                                                           ' ' ' -                                          ,          - - - - - -                 . . - -_             y_ . ,, _ _ _ _ _ _ . ,

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also worthwhile to note that the highly reliable (redundant) AC-independent decay heat removal systems provide added value when ability to cope for long durations exists and very low core melt frequencies are estimated. Figure 8.2 shows the sensitivity of various emergency diesel generator reliability levels on estimated core damage frequency. A combination of reasonably good diesel generator reliability and ability - to cope with a several hour station blackout results in estimated core 2 -- damage frequencies on the order of 10-5 per year or less. The effect of h emergency AC power configuration is shown in Figure 8.3. Note that . - -U;2 _- + 1 ._ Y substantialdifferenceincoredamagefrequencymayexistbetweenplhints E-hh with three emergency diesel generators depending on the minimum number

(1 or 2) needed to maintain core cooling and decay heat removal during a loss of offsite power. Again, frequencies drop rapidly as station blackout coping capabilities extend to cover longer duration AC power outages. The last sensitivity provided in Figure 8.4 shows results for variations in emergency diesel generator failure rate from both independent and cccmon causes. Common cause failure in support systems __ QJ (e.g., service water, DC power) have been held at ncminal values. Theseg.=i : =]=-g - results shcw the potential for low estimated core damage frequency by i maintaining highly reliably emergency AC power systems. Core damage -

,           frequencies on the order of 10-6 per year may be possible if a high state of operational reliability can be maintained in the emergency AC        .   .
                                                                                                                         ]{

power system combined with some capability of coping w'ith an unlikely-

                                                                                                                                ~ ~ ~]

station blackout. m j t. __. 7=:- -

~ =-

ify- - g i- p 2 :t g r_ The preceding results and additional sensitivity analyses can be usedjo_ y dfL5_ . _ _ __.2 assess the effectiveness of certain strategies in dealing with stationEI 5

blackout concerns. For instance, if in a PWR the reactor coolant pump seals were known to fail early during station blackout, core damage could occur in one or two hours after loss of AC power even if the AC-independent decay heat removal system (AFWS) were operating properly. l Table 8.1 has been developed from the sensitivity analyses to show the effect of providing a "fix" to maintain reactor coolant pump seal i i NUREG-1032 .- .-- .

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i Station Blackout Capability ("'0"#5} Figure 8.2. Sensitivity of Estimated Station Blackout-Core Damage Frequency to EDG Reifability, AC-Independent Decay Heat Removal Reifability, and S,tation Blackout Coping Capability i NUREG-1032 - 74 - F g ..wr_._. , .._._y .._.t---3-m-_.,.:7..,,._.,--_,,-.v.._-.,_._, _.b

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10'8 8 12 16 0 4 Station Blackout Cacability (Heurs) Figure 8.3. Sensitivity of Estimated Station Blackout-Core Damage Frequency to Emergency AC Power Configuration, AC-Independent Decay Heat Removal Reliability, and Station Blackout Coping Capability I t 75 - T l NUREG-1032 p

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4 8 12 16 Station 51ackout Capability (Hours) Figure 8.4 Sensitivity of Estimated Station Blackout-Core Damage Frequency to Reducing EDG Comon Cause Failure Susceptibility, EDG Reliability, and Station Blackout Coping Capability NUREG-1032 .

Table 8.1 Sensitivity of Estimated Core Damage Frequency Reduction for Station Blackout Accidents with Reactor Coolant Pump Seal Failure Delay frcm 2 to 4 hours and 4 to 8 hours __. z . 1/2 Configuration . .

 =.==                                                                                                                             w= = .                          . . .

N5$$ EDGR = 0.025 EDGR = 0.05 55 5_~ l ' _ . 5 ' ~ 2 "

   ---~' Cluster                                                      2 to 4 hr        4 to 8 hr             2 to 4^hr      4 to 8Thr2~-                Mri
                                                                   - 2.8E-6            1.2E-6                6.2E-6'        2.5E-6

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Cluster 2 to 4 hr 4 to 8 hr 2 8.3E-7 3.4E-7

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integrity to allow successful core cooling for station blackouts of four

           . and eight hours.

The results provided up to this time represent point estimates of probability or, more properly, frequency. The effect on the mean probability estimates of using log normal distributions to represent basic event probabilities, .,- calculated medians and uncertainty rang'es was shown in. Reference 5 _When that

              - work wa~s completed, the magnitude of the uncertainty in-thet loss-ofr. offsi te -:-
 '~ 7
               ' power ~ frequency and duration estimates was not known.:: Because the uncertainty -       e;-             3-2;;-

bounds a're now perceived to exceed those used in that work, the accident. -

sequence uncertainty ranges derived using the most recent uncertainty estimates for loss of offsite power frequency may be larger than previously estimated. The loss of offsite pcwer frequency and duration estimates are most uncertain for the very low frequency long duration losses of offsite power. The uncertainty on the probability of accident sequences which result from the shorter duration losses of offsite power should not be significantly different from the previous estimates. Some typical station blackout core damage probabilities and uncertainty ranges representing a 90% confidence interval have been provided in Figure 8.5 for reference. It can be seen that the sequence mean is typically 3 to 8 times larger than the point estimate and that the upper and lower bounds are

    ~                                                                                                                   -

l

              ~ typically within a factor of 5 to 20 of the median estimate. The large .                            .
         ~

difference in point estimate and mean can be attributed-to-the use of a-log - .: normal distribution. When sequences are combined into a single core damage . -;.- probability, the proportional distance between mean and point estimate tends to decrease somewhat. A measure of risk associated with station blackout accidents can be obtained by multiplying the estinated core damage likelihood by the estimated dose due to containment failure during a station blackout accident. The recovery of AC power during the accident would provide the potential for terminating core damage prior to core melt and the potential for reducing fission product releases by delaying containment failure or actuation of containment sprays NUREG-1032 l

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I Figure 8.5. Estimate Core Damage Frequency Showing Uncertainty Range for Four Reference Plants from NUREG/CR-3226 NUREG-1032

                                                                                                                                  -   79 -

9 l l prior to containment failure. Some perspectives on estimated risk are provided in Appendix C.

9. Relationship of Other Safety Issues to Station Blackout There are several safety issues for which station blackout implications were .

reviewed for significance. These include: loss-of-coolant accident initiators; anticipated transients without scram; external. hazards, such as

         ' seismic events and severe weather; and internal hazards associated with fire         -       -

or extreme envirorments, such as flooding or high steam temperature due to pipe breaks within the plant. In general, it was concluded that if station blackout likelihood were independent of any of these other safety , considerations, the potential risk of a station blackout concurrent with one of those other safety concerns is very small. However, if due to common cause failure or interactive failure the initiation of an accident by one of those other mechanisms described causes a station blackout, then the safety implications of those safety issues on station blackout are fairly large. Each of these safety issues is discussed separately below. (a) Loss-of-Coolant Accidents _ _ _ . _ . ~ Loss-of-coolant accidents induced by a station blackout transient have already

been included in the accident sequence analyses described in Section 7. These

        ~ will not be discussed further here. Loss-of-coolant accidents concurrent with.

a loss of offsite power are usually included in the design basis of nuclear - power plants in accordance with the general design criteria of Appendix A to 10 CFR 50. The likelihood of a loss of coolant accident followed by and concurrent with a station blackout has been considered and is discussed below. Although no strong coupling could be found between the initiation of a loss-of-ccolant accident and a subsequent failure of the offsite or onsite AC power system, one potential mechanism has been identified. If a loss-of-coolant accident were to occur at a nuclear power plant, the reactor would trip following which the turbine-generator would be tripped and a grid instability could follow, or the site could be isolated by switching activities in the NUREG-1032 switchyard to provide onsite safety-related or alternative sources of preferred I power to the emergency power safety buses. Historical experience collected on loss-of-offsite-power events at nuclear power plants suggests that the likelihood of a failure of the offsite power supply, given a transient or an l accident situation which would cause a trip of the turbine-generator is on the

         ~~ order of 10-4 to 10-2 , depending on the strength of the grid and the offsite                                                          .
,               power design at the site.
       ' E ~-      --      -

[I '5stimated' loss-of-coolantaccidertfrequenciesrangefrom10-2,per reactor-year - - r-

              'for small loss-of-coolant accidents down less than 10-4 per reactor year for                                               -

i large diameter pipe breaks. The small loss-of-coolant accident frequency is dominated by pump seal LOCAs on pWRs and stuck open SRVs on BWRs, situations i which do not require rapid actuation of AC-powered emergency safety feature ! equipment and which have been addressed previously. The most likely small loss-of-coolant accident which has not been incorporated in the station blackout accident analyses is a small pipe break (less than 2" dia.) with a ! frequency of about 10-3 per reactor-year.

The icw frequency of loss-of-coolant accidents combined with the likelihood of losing offsite power on turbine-generator trip results in an estimated ~ '~

frequency of occurrence ranging from 10-5 per reactor-year to 10-7 per I~ ~ ~ reactor-year. When combined with a conservative estimate. of. emergency A0 power - ! systen unreliability of 10-2 per demand, it is easily.shown that. accident . - -

    ~     ~

iequences of this type represent a small element of reactor risk (less than- --

  ~
               ~ 10-7 perreactor-year). The variability of the frequency of station blackout--                                                       -!

l caused by a loss-of-coolant accident could be as much:as two orders of - i ! magnitude higher and still represent one of the smaller station blackout accident threats. Although, at this higher level, these accidents could i represent a noticeable fraction of reactor risk. Large pipe break less-of- l coolant accidents with initiating frequencies on the order of 10-4 per ( i l reactor-year combined with the probability of subsequent failure of all AC power do not appear to represent an appreciable fraction of accident likelihood or public risk, at least in comparison to other station blackout sequences. NUREG-1032 ! f ! l.._------.-.. _ _ _ _ _

         ~ - . - - _ _                   _ - . _      _    _ _ _ _ . . _       _ -         . _ _ - . _ _                  _ _ - . - _ - - _                  __     - - - _ _ _ _ _ - - -

l (b) Anticipated Transients Without Scram ! . l j Another safety consideration that was investigated is anticipated transients ' I without scram. In this case, the anticipated transient is a loss of offsite power. If the probability of a loss of offsite power is taken as the generic l i'

        - average, 0.1 per year, and the probability of reactor. scram failure is taken as I            the historical average, about 10-4 per demand, then the probability of a loss                                                                                                                     .'        [

I

!-          U offsite power followed by a failure to scram is about.105.. .This is a level;                                                                                                     .,

_y ? ~ ~ W accident sequence likelihood that might be considered. important. . However,- ...

                                                                                                                                                                                                    ....'-.l C            in order for station blackout to occur, the onsite emergency AC power system                                                                                        - .. ..                            .
;           must also fail. In the worst case, one might find an unreliability of the emergency AC power system of about 10-2 per demand.                                            Thus, the frequency of an                                                                                    i anticipated transient without scram involving loss of offsite power and a                                                                                                                                  ;

I failure of the onsite emergency AC pcwer system is on the order of 10 per reactor-year. Even ifethe level of uncertainty.were an order of magnitude . l f ! higher, this accident sequence would not be of concern in comparison to the - dominant station blackout accident sequences that have been identified. i i l (c) Extreme Internal Environment 4 ! A safety area in which there does appear to be a potential for station . N blackout type accident sequences being induced by other.causes involves fire g . F and other extreme environments internal to a nuclear power plant.- .The ..; - l' concern associated with internal environmental hazards.is that their. . . . . . . . :. y i occurrence can represent a comon cause accident initiator which also.affects .--... ..I the ability to cope with the incident. Specifically of concern is the -

likelihood of a fire, flood, or other extreme environmental condition __ generated by internal events which would cause a loss of all AC power. In general, in order for this to occur there must be portions of AC power i systems in a connon location where these hazards are present; or protection barriers and AC power system design requirements must be insufficient to ; i control the spread or failure due to these hazards. Therefore, the . l t L t i NUREG-1032 j

   ,y      _ ,._.._... _ ._., _. ,_,.... = ,__ -._.,,_., , -. _._._ , _, _ _ .:._ _ , . , , _'

__ _ , . , _ _ , . .s___. , , _ _ _ , , . . _ . _ . . , _ _ _ - _ - - , -

i likelihood of internal hazards causing a station blackout type of accident is heavily dependent on the plant's design and in particular on the location of equipment. If separation and internal environmental protection barriers are maintained, or adequate AC system design is provided, the likelihood of these internal environmental hazards causing a station blackout type of accident would be very small, probably less than 10'0 per reactor-year. On the , other hand, if comonality of location or a lack of protection exists at

F ~ a plant, then the safety significance of these internal hazards would .. . . . .

        '~~2 need to be evaluated for plant damage susceptibility.. The frequency of;      .    ..

7 occurrence of these hazards can be as high as once per one.hundred to . once per one thousand reactor-years. Therefore, the vulnerability to station blackout due to these hazards can be of concern. . 1 (d) External Hazards Another potentially significant safety consideration that could be related to station blackout involves external hazards to the plant, particularly those due to seismic and weather-induced failures. To date, a seismically induced loss of offsite power has not been observed at a nuclear pcwer plant, but such events have occurred at fossil-fueled generating stations. Failure of offsite power due to severe weather has been observed at nuclear power plants. In fact, severe weather was included as a major factor in,

, determining the likely duration of an extended offsite power outage at nuclear power plants as described in Section 3. The greatest potential.for. . .. . safety significance exists where there is a direct coupling or comon cause failure associated between a transient initiating external hazard causing loss I~~ of offsite power and the reliability of the onsite and offsite power systems. ! It can be expected that significant seismic and severe weather events will < cause a loss of the offsite power system. On the other hand, the plant, and in particular the emergency AC power system, is typically designed to withstand, or is protected from the effects of, these severe phenomena. Therefore, for severe external hazards that are within the design basis of the plant, the tailure of the emergency AC power system can be considered as an independent NUREG-1032 .- .. ._ .

f e 9 failure event. For example, if the likelihood of a safe shutdown earthquake which could cause a loss of offsite power yere approximately 10-3 per year or ,,, less, and one assumes that it would take approximately 8 to 24 hours to restore offsite power from such an incident, then a typical estimate of core damage or core melt frequency due to a safe shutdown earthquake and a station blackout would be about 10-6 per reactor-year or less. For severe weather the . likelihood of the weather-induced failure of the offsite power system could be.

                              ~
     "2 as high as 10-2 per year, and the outage could be expected.to.be- on the order                ;
                                                                                                                     . 3
         'Iof'16' hours or more. Again, if the severe weather event-is within.the. design;
                        ~

basis of the plant, the likelihood of a weather-induced. station. blackout .. accident causing core damage or core melt would be on the order of 10-5 per . reactor-year. Table 9.1 provides a summary of the typical internal and external accident hazards of a nuclear power plant and an identification of some potential

   ~
                     ~ points of failure that could result in a coupling between.these. accident                             .

initiators and a station blackout. If such interactions or. points of

commonality do not exist, then it is concluded that the_ contribution of ., these accident initiators to station blackout accident sequences results in core melt are no larger, and probably much less, than those previously l considered. O i 0 NUREG-1032

r I' . , I l Table 9.1. Coupling Between External (and Internal) Events and Potential Plant Failures

i - - ----- --- - - EV ENT POTENTIAL PLANT " WEAKNESS" Seismic Switchyard, Control, Non-Seismicly Designed Equipment etre, Flood Areas with Multiple Divisions, Inadequate Protection Barriers Severe Weather Transmission Lines and Towers, Switchyard, Non-Safety Structures 3 e NUREG-1032 I i References .. .

1. U. S. Nuclear Regulatory Commission, " Reactor Safety Study," NRC Report WASH-1400, NTIS, October 1975.

2. H. 'Wycoif, "Lesses of Offsite Power at U. S. Nuclear Power Plants-All ,

Years Through 1983," NSAC/80, EPRI, May 1984.

3. R. E. Battle, " Collection and Evaluation of Complete and Partial Losses of Offsite Power at Nuclear Power Plants," NUREG/CR-3992, to be published.

4. R. E. Battle and D. J. Campbell, " Reliability of Emergency AC Power Systems at Nuclear Power Plants," NUREG/CR-2989. July 1983.

5. A. M. Kolaczkowski and A. C. Payne, Jr. , " Station Blackout Accident Analyses (Part of NRC Task Action Plan A-44)," NUREG/CR-3226, May 1983.

6. U. S. Nuclear Regulatory Commission, " Clarification of TMI Action Plan Recuirements," NUREG-0737, November 1980.

7. C. D. Fletcher, "A Revised Summary of PWR Loss of Offsite Power Calculations," EGG-CAAD-5553, September 1981.

8. J. P. Adams, et al., " Natural Circulation Cooling Characteristics During

' PWR Accident Simulations," Second National Topical Meeting on Nuclear , Reactor Thermal Hydraulics, January 11-14, 1983.

9. F. E. Haskin, W. B. Murfin, J. B. Rivard, and J. L. Darby, " Analysis of a Hypothetical Core Meltdown Accident Initiated by Loss of Offsite Power for the Zion 1 Pressurized Water Reactor," NUREG/CR-1988, December 1981.

NUREG-1032 j

10. R. R. Schultz and S. R. Wagoner, "The Station Blackout Transient at the Browns Ferry Unit One Plant A Severe Accident Sequence Analysis,"

EGG-NTAP-6002, September 1982.

;   11. D. H. Cook S. R. Greene, R. M. Herrington, S. A. Hodge, and D. D. Yue,
          " Station Blackout at Browns Ferry Unit One - Accident Sequence Analysis,"                                                       ,

NUREG/CR-2182, November 1981.

12. NUREG-900, " Nuclear Power Plant Severe Accident Research," January 1983.

13. IDCOR Technical Sumary Report, " Nuclear Power Plant Response to Severe Accidents," November 1984.

e I ' 1 i l i

[ i i l i l NUREG-1032 t g e g. o,a *P

a , 1 - , , - , - - _ . - . - , - - - - - - - -

DRAF Appendix A. Development of Loss of Offsite Power Frequency and Duration Relationships Introduction _. This Appendix provides the details and results of analyses performed to develop the cause, frequency, and duration relationships for loss of offsite power at nuclear power plants. The purpose of this work was to develop generic less of offsite power relationships which would allow differentiation of plant design, operational, and location factors which can significantly affect the expected frequency and duration of offsite power losses. Within this study, the loss of offsite power has been defined as the interruption of the preferred power supply to the essential and' nonessential switchgear buses necessitating or resulting in the use of emergency AC power supplies. A total loss of offsite power is said to have occurred when non-emergency AC power sources become unavailable requiring some diagnosis or special recovery cctions including correcting switching errors, fixing or bypassing faulted cquipment or otherwise making available an alternato standby source of non emergency AC power. Although total loss of offsite power is a relatively infrequent occurrence at nuclear power plants, it has happened a number of times in the past, and a data base of information has been compiled from past experience (References 1 and 2). From these data and a review of relevant design and operational characteristics, the frequency and duration relationships for loss-of-offsite power events at nuclear power plants have been developed. Historically, a loss of offsite power has occured with a frequency of about once per ten site years. The ! typical duration of these events has been on the order of one-half hour. However, there has been experience at some power plants in which the frequency f of offsite power loss has been substantially in excess of the average, and in [ other instances the duration of offsite power outages has greatly exceeded the { , norm. NUREG-1032 Appendix A A-1 ,

e. - - - - , . .-. . . , - , - _ , _ . - - . . . - . r . -- -, , , ,- ,------- --

a Table A.I. Summary of Loss of Offsite Power Experience No. of Events Frequency Per _. Category (> hr) Site-Year (>h hr) _ ,_, Plant Centered 30 0.056 (15) (0.020) Grid 10 0.019 - (4) (0.007) Weather 6 0.011 (5) (0.009) Total 46 0.086 (24) (0.045) , Interruptions , 8 0.015 (0) (0) Total + Interruptions 54 0.100 (24) (0.045) a Data through 1983 excluding sites with one offsite power connection (Humbolt . Bay, Lacrosse, and Big Rock Point prior to March 1968). Number of site years through December 1983 is 533. NUREG-1032 Appendix A A-2

A summary of the data on the total loss of offsite power events is provided in Table A.1. Since design characteristics, operational features, and the location of nuclear power plants within different grids and meteorological areas can have a significant effect on the likelihood and duration of loss of offsite power events, it was necessary to analyze the nuclear industry experience in more detail. The data have been categorized into plant-centered events and , area or weather-related events. Plant-centered events are those in which the design and operational characteristics of the plant itself play a role in the likelihood or duration of the loss of offsite power event. Area or weather effects include the reliability of the grid and external influences on the grid or at the site, such as severe weather, which have an effect on the likelihood end duration of the loss of offsite power. The data show that plant-centered svents account for the majority of the loss of offsite power occurrences. The trea blackout and weather-related events, although of lesser frequency, typically recount for the longer duration outages with storms being the major contributor to long duration outages. - Figure A.1 provides a plot of the frequency and. duration of loss-of-offsite-power events due to plant-centered faults, grid blackout, and severe weather, based on past experience at nuclear plant sites. The curves were developed by fitting data to a two parameter Weibull function of the following form:

                                                         . (4; lt b)

A to r; (0 - A L o P s. e where ALOPi(t) is the frequency of losses of offsite power of type "i" which are cqual to or greater than duration "t". That is, the recovery time equals or exceeds "t" hours. The term A is the frequency of occurrence of offsite LOPi power losses of type i which have greater than zero duration, and parameters aj l cnd pg are curve shaping constants which very according to the data being curve fitted. I j NUREG-1032 Appendix A A-3

4 Figure A.1. Loss of Offsite Power Frequency vs. Duration I. di!l,...' . af. i?t'

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Loss af Offsite Power from Plant-Centered Causes Plant-centered failures typically Iinvolve hardware failures, design deficiencies, human errors (maintenance and switching), and localized weather-induced faults (lightning), or combinations of these failure types. Plant-centered failures recoverable by switching or repairing faulted equipment at the site. An attempt was made to determine if any correlations existed between offsite power design characteristics and frequency and duration of losses of offsite power for the plant-centered losses. Two offsite power design features were identified as being potentially significant with regard to loss of offsite power frequency and duration. These features included independence of incoming offsite power sources and the number of immediate and delayed access circuits and their transfer schemes to the class 1E buses. Table A.2 provides a definition of design differences associated with these features. The designs for offsite power sources were further subdivided into groups, and the number of shutdcwn sources were subdivided into different possible design combinations (Reference 2). , The relationship between the listed design features and loss of offsite power frequency was analyzed using the Failure Rate Analysis Code (FRAC) (Reference 3) to identify a loss of offsite powar frequency correlation for various design features. Results of these analyses revealed no statistically significant correlations involving frequency of plant-centered loss of offsite power and the design features analyzed. An analysis was also performed to determine if a statistically significant relationship exists between offsite power design characteristics and loss of offsite power duration. An analysis of covariance was performed to check for a relationship between frequency and duration using the generalized linear model (GLM) procedure of the Statistical Analysis System (SAS) (Reference 4). The NUREG-1032 Appendix A A-5

Table A.2 Definitions of Offsite Power System Design Factors A - Independence of offsite power sources to the nuclear power plant

1. All offsite power sources are connected to the plant through one switchyard ,,

2. All offsite power sources are connected to the plant through two or more switchyards, and the switchyards are electrically connected.

i

3. All offsite power sources are connected to the plant through two or more switchyards or separate incoming transmission lines, but at least one of the ac sources is electrically independent of the others.

B - Automatic and manual transfer schemes for the Class 1E buses when the normal source of ac power fails and when the backup sources of offsite power fail i . .

1. If the normal source of ac power fails, there are no automatic transfers and one or more manual transfers to preferred or alternate offsite power sources.

2. If the normal source of ac power fails, there is one automatic transfer but no manual transfers to preferred or alternate offsite power sources. ,

a. All of the C1. ass lE buses in a unit are connected to the same ,

preferred power source after the automatic transfer of power sources. J

b. The Class 1E buses in a unit are connected to separate offsite power sources after the automatic transfer of power sources.

i NUREG-1032 Appendix A A-6

Table A.2 Definitions of Offsite Power System Design Factors (continued)

3. After loss of the normal ac power source, there is one automatic transfer.

If this source fails, there may be one or more manual transfers of power sources to preferred or alternate offsite power sources. .

a. All of the Class 1E buses in a unit are connected to one preferred power source after the first automatic transfer.

b. The Class IE buses in a unit are connected to separate offsite power sources after the first automatic transfer.

4. If the normal source of ac power fails, there is an automatic transfer to a preferred source of power. If this preferred source of power fails, there is an automatic transfer to another 50urce of offsite power.

a. All of the Class 1E buses in a unit are connected to the same preferred power source after the first automatic transfer,

b. The Class 1E buses in a unit are connected to separate offsite power sources after the first automatic transfer of power sources.

NUREG-1032 Appendix A A-7

l Type IV Sum of Squares was used for all calculations. No statistically significant relationship between frequency and duration was found. Thus, no additional covariance analyses were run. Substantially, the data for all of I the different design factors were analyzed to check for any statistical interactions using analysis of variance. One data point, a 5.83 hour restoration time for an event at Calvert Cliffs on 4/13/78, was found to cause , a strong intaraction. Without that event there was no significant interaction. That event involved a latent design flaw which has since been corrected and is not expected to typify future occurrences with regard to design feature, type - of failure, and duration. With the data " corrected," both types of design features were found to be statistically significant determinants of plant-centered loss of offsite power restoration time. Thus any combination of the design features analyzed could be used to define a set of design characteristics with a statistically different recovery time for plant-centered losses of offsite power. From this analysis it was concluded that plants with switchyard designs which are normally operated as an interconnected system could be separated as a group from designs which offer electrical independence, and that sites with two or more alternate offsite power circuits (immediate or delayed , rccess) in addition to the normally energized power circuit to the 1E buses (offsite or unit generator source) could be grouped. Table A.3 shows design combinations thus obtained with the mean time to repair (MTTR) values for each group and the statistical test values which support this grouping. It should be noted that other groupings can be derived which are both statistically significant and physically valid. However, data limitations and , small differences in MTTR which occur for more detailed breakdowns suggests that the design groups obtained represent a reasonable and valid compromise between

completely generic and more design-specific breakdowns. l A plant-centered loss of offsite power frequency vs. duration curve was l developed for each of the four design groups by fitting the corresponding data to a two parameter Weibull distribution. A list of the data used for each curve fit is given in Table A.4. The actual curves generated by this analysis t are presented in Figure A.2. Figure A.2b shows the 90% confidence limits for two of the correlations (Il and I4) derived using the extreme value theory. NUREG-1032 Appendix A A-8

Table A.3 Mean Time to, Restore Offsite Power and Statistical Test Values for Plant Design Groupings Group Design Mean Time to Restore Designation Features

Offsite Power (hrs) _ _ . , ,

11 A3 and (B3 or 84) 0.15 f I2 A3 and (B1 or B2) 0.29 f r I3 (Al or A2) and (B3 or B4) 0.44 f I I4 (Al or A2) and - (B1 or B2) 1.08 ; f , Statistical Test Values

1. F value Pr F L A 7.77 0.0094 8 3.93 0.0573 A*B 1.61 0.2150 1

         *A1, A2, A2, B1, B2, B3 and B4 are defined in Table A.2.                                                                       I i

l Note: Frequency of plant-centered loss-of-offsite power events was 0.056 per [ i site year. ; i I i' I i t t t NUREG-1032 Appendix A A-9

Table A.4 DataUsedforPlant'CenteredgossofOffsite i Power Duration Curve Fits Site Date Duration (hours) Group II: Fitzpatrick 10/04/78 0.004 b Oconee 01/04/74 0.013 Fitzpatrick 03/27/79 0.05 Millstone 07/21/76 0.08 C ' Indian Point 2,3 06/03/80 0.50 Group 12: Nine Mile Point 11/17/73 0.003 Haddam Neck 07/19/72 0.01g - Haddam Neck 07/15/69 0.15 - Haddam Neck 06/26/76 0.27 Haddam Neck 08/19/74 0.33 Haddam Neck 04/27/68 0.48 , Group 13: Davis Besse 11/29/77 0.002 Oyster Creek 09/08/73 0.00j Point Beach 04/27/74 0.02 . Brunswick 2 03/25/75 0.07 Monticello 04/27/81 0.25 Beaver Valley 07/28/78 0.28 Davis Besse 10/15/79 0.43 Ginna 03/14/71 0.50 Quad Cities 06/22/82 . 0.57 Ginna 10/21/73 0.67 Prairie Island 07/15/80 1.03 Quad Cities 11/06/77 1.15 Arkansas-1 09/16/78 1.48 Group I4: San Onofre 11/22/80 0.004 , Fort Calhoun 08/22/77 0.015

  .                       Palisades            09/24/77                  0.50 Farley               09/16/77                  0.90 Fort Calhoun         02/21/76                  0.90 Palisades            09/02/71                  0.93 Indian Point         06/03/80                   1.75 Farley                10/08/83                 2.75 a Not      included in the duration analysis were the Palisades events of 11/25/77                              '

and 12/11/77 (recurring failures), the Calvert Cliffs event of 04/13/78 , (outlier), the Big Rock Point event of 11/25/72 (insufficient plant design 1 information), and the Crystal River event of 06/16/81, the Vermont Yankee event of 12/17/72 and the Turkey Point event of 04/04/79 (incomplete reporting of event durations). b For events with unspecified durations of less than one minute, durations were assigned to facilitate the statistical analysis, c The Indian Point event of 06/03/80 lasting 1:45, included in Group 13, is included as a 0:30 event in Group Il on the basis that had the available gas turbine been employe'd, offsite power would most likely have been recovered , in approximately 30 minutes. A-10

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It was recognized that some of the loss-of-offsite power events represented

       " infant mortality", or burn in, and as experience is gained aiid problems fixed the expected frequency could drop. Figbre A.3 is a plot sho, wing the actual cnd median smoothed plot of time between failures (loss of offsite power events) as a function of the event number. The trend appears to show an increase in time between failures, or decrease in frequency. Visual inspection    ,

of the median smoothed time between failures plot indicates a reasonable break point at about the 7th occurrence, which roughly corresponds to January 1978. Based on an analysis of variance, it was concluded that the loss of offsite power frequency due to plant-centered events from 1978-1983 decreased by 30 percent from what it was previously, for events lasting one-half hour or longer. It should be noted, however, that with the removal of the occurrence of just one event in 1977, statistical support in this trend would drop substantially. Another test performed to evaluate the learning ' factor was to see if any corre-lation exists between plant age and loss of offsite power frequency. Visual examination of the data indicated a rather random frequency of occurrence based on plant age. It was concluded that there probably has been some improvement in the loss of offsite power frequency for plant-centered events at nuclear power plants as operating experience has increased, but that some additional time and evaluation will be needed to definitively show the permanence of such an observation. Loss of offsite power frequency estimates provided later in this Appendix are based on the reduced frequency of plant-centered events (0.04'per site year) obtained as a current best estimate. 1 i l l [ 1 NUREG-1032 Appendix A A-13 l l l

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Grid-Related Loss of Offsite Power Grid reliability has traditionally been the most prominant factor associated with a loss of offsite power at nuclear power plants. Yet, the historical data show that losses of offsite power due to grid-related problems account for no more than 20 percent of all losses of offsite power. Attempts to find , characteristics to classify site, design and location features which affect the cxpected frequency of grid loss have not been successful. An investigation into the various utility transmission and distribution system reliability characteristics was beyond the scope of this study and would likely involve an extensive state-of-the-art analysis of grid stability, the results of which would be of questionable validity considering liuiitations on current methodology. In its place a more pragmatic and experience-based approach to cstimating nuclear plant site susceptibility to grid loss was taken. Both frequency of grid loss and time to restore power were considered. It was recognized that the Florida Power and Light (FPL) grid has represented t.M a upper end of utility grid failure frequency during the past 10-15 years, " c1though some recent improvements seem to have been effective. Very few other nuclear plant sites have experienced even one or two loss-of-offsite power cvents due to grid blackout. The great majority of nuclear power plants have not experienced grid failure. A systemic weakness which has been identified following the occurrance of a grid failure is usually corrected as soon as possible. As such, it is usually a new and unidentified systemic weakness which results in future failures. Therefore, in the absence of known and uncorrected systemic weaknesses, the occasional, non-recurring type of grid failure may not be a good indicator of future trends within a utility system. With this in mind, the FPL experience was separated from the balance of the U.S. nuclear utility experience to estimate grid failure frequency. Since a set of design or location factors could not be identified which could NUREG-1032 Appendix A A-15

effectively differentiate the expected grid reliability of the various utility systems, grid reliability was categorized into failure frequency ranges characteristic of past experience. The FPL experience suggests an upper end to j the grid failure frequency of once per two to four site years, although there i have been recent 1mprovements. In other utility systems, the occasional grid i failure appears to have a frequency of about once per ten site years, and the rare or unexpected grid failure frequency can be characterized as occurring Icss than once per twenty site years with a national average of' once per one hundred site years, excluding FPL experience. Table A.5 provides a list of grid-related losses of offsite power and site-specific frequencies calculated from the data. Two grid undervoltage events are also shown in the data. 4 Although not counted as grid failures, offsite power sources were momentarily unavailable during these events. There are two factors which have been identified as significant in determining the duration of grid-related losses of offsite power at nuclear power plant sites. .These are the availability of adequate restoration procedures and the availability of " black start" power sources able to supply power to a nuclear power plant in isolation of a grid disturbance. Both of these factors can

contribute to a significant reduction in the expected duration of grid-related loss of offsite power as reported in the Indian Point Safety Study (Reference 5). In 1981 the NRC sent a generic letter (Reference 6) to all nuclear power plant licensees requesting them to develop and implement procedures to enhance 4

restoration of offsite power. Responses to that generic letter have indicated that at many sites power could be preferentially restored to a nuclear power , plant within one or two hours, even if the grid remained in a blackout condition. , i I NUREG-1032 Appendix A A-16 l - _ _ . , _ _ . __ .- ___ -- . .. - - -- _

6 'I J . i Table A.5. Grid-Related Loss of Offsite Pcwer Frequency , Versus Duration Data Through December 1983 Date of Site Frequency - _ Site Occurrence Duration (hours) (per year) _ ) Turkey Point 04/03/73 0.30 . 04/04/73 0.25 t 04/25n4 0.33 , 06/28n4 8 0.18 1.03 . 05/16/77* 2.00 0.446 (5 events in i 05/16/77 ] 11.4 site years) i i 4 Indian Point 11/19/65 ? 07/20/72 0.92 0.15 (3 events in , 20 site years) 07/13/77 6.47 D ' St. Lucid 05/16/77 D 0.33 j 05/16/77 1.50

 ,                                              05/14/78                                                      0.13                                                            0.260 (2 events in 4                                                                                                                                                                              7.7 site years) l               Yankee Rowe                      11/19/65                                                      0.65                                                            0.044 (1 event in                                                      ;

4 22.5 site years) c I l 45 si.tes none -- 0 (no events in ' l 3.5 to 23.4 site-years) l Total for all 50 sites 0.019 (10 events in l (539 site years) ! Total for Sites Excluding FPL 0.008 (4 events in ' (530 site years) j l "The Turkey Point events of 05/16/77 were counted as one event for frequency calculations. , ; j bThe St. Lucia events of 05/16/77 were counted as one event for frequency l ] calculations. i "The undervoltage event at Millstone on 07/21/76 was treated as a plant-centered ! design problem; the undervoltage event at Quad Cities on 02/13/78 was treated : j as a degredation with a useable offsite power source available throughout i the incident. ; t I-i j i NUREG-1032 Appendix A A-17 ! l , I l

   ,,   ..-c    -.,-.,w_---   .--.,.,_w..--ro--       ,.~.,,,--._m.~,m._                 _, .,, _ _ , ,      , , , , , , , , , . . . . ,     ,   , . , _ _ , , , _ _ _ , , , , ,   ,_,,,,__ ,_ _. _ _ , , , - , _         _ _ , _ , ,         _

a . l l The offsite power restoration time during grid failure can be estimated by past

cxperience, or if an appropriate set of procedures are provided and power I sources are available and capable of supplying power during grid blackout, a ! acre prompt recovery may be possible. Human reliability and alternate power source availability will limit the potential for recovery to as low as 60 percent recovery in about an hour or it may be as high as 95 percent recovery ,

in less than one-half hour if multiple, reliable sources of power which are isolatable from a blacked out grid are available. For this study an offsite

, power restoration likelihood of 80 percent within one-half hour of a grid failure was assumed for analysis of plant sites with enhanced recovery (e.g., procedures and power sources available for prompt recovery). Grid-related l loss-of-offsite power probability of recovery was developed by fitting past , l operating data to a two parameter Weibull distribution. The data used in the curve fit are provided in Table A.5. Figure A.4 provides a curve showing the probabi' . / of non-recovery versus duration for losses of offsite power due to i grid bla.. outs and shows the potential for improvement of enhanced recovery I capability over past operating experience. j . l The grid reliability and offsite power restoration correlations were developed by combining the occurrence frequencies representative of those observed in J past operating experience and the calculated recovery probabilities. Table A.6 provides the grid failure frequency and duration groups obtained. Figure A.5 shows the discrete loss-of-offsite power frequency and duration curves corre-l

sponding to the groups identified in Table A.6.

Loss of Offsite Power Due to Severe Weather i

l Severe weather conditions, such as local or area wide storms, have caused , losses of offsite power at nuclear power plants. Weather-related causes of offsite power failure have been divided into two groups: i l NUREG-1032 Appendix A A-18

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, < 1 per 20 site-years (.. 1/yal G3 > 1 per 20 site-yeen and i 1 pe' E site-years (o.2/ye) G4 Greater thar. or equal to 1 per 6 site-years (o s/ye ) I . Recovery (R) -
Recovery fem Grid Recovery Capability Blackout Groups (R)
R1 Plant has capability and roce p(nonemergency) dures to recover AC power offsite to the site within 1/2 hour - j following a grid blackout. { R2 All other plants not ir. RI. Grid Reliability / Recovery (GR) . Grid Reliability / Recovery Grid Reliability Recovery 'ror. Grid Group (GR) Group (G) Blackout Groue (R) GR1 G1 R1 GR2 G2 R1 GR3 G3 R1 GR4 G4 R1 GR5 - G1 R2 GR6 62 R2
GR7 , G3 - R2 J
j NUREG-1032 Appendix A A-20 _ _ _ . ~ - _ _ _ _ _ . . - . _ _ . _ __ _ - _._ _ . . _ _ _ _ _ . . _ - _ _ _ . . . _ _
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Figure A.5. Grid-Related Loss of Offsite Power Frequency and Ouration Groups NUREG-1032 Appendix A A-21 a
(1) Those for which the weather caused the event but did not affect the time to restore power; and (2) Those for which the weather initiated the event and also caused conditions such that power was not or could not have been restored for a long time. The former group includes lightning and most other weather events that do not cause severe or extensive physical damage at or near the site. They can cause a loss-of-offsite power, but their severity does not contribute in any signifi- - cant way to long duration offsite power losses. These types of weather-related Gffsite power outages are usually treated in the plant-centered or possibly grid category. The latter group includes losses of offsite power due to major storms, hurricanes, high winds, snow and ice accumulations, and tornadoes. The loss of cffsite power expectation frequency of this group is relatively small. On the other hand, the likelihood of restoring offsite power in a short time from this group is also relatively small. In order to estimate the likelihood and duration of loss of offsite power due to severe weather it is necessary to identify the set of weather hazards to be considered, deterr.ine the likelihood of failure for a given hazard intensity, cnd determine the repair or restoration time for the various failure modes associated with severe weather-related power losses. Although utilities and regional power pools normally keep extensive data on transmission line, terminal, and customer outages due to all causes including weather, little information has been obtainable which can be used to derive the likelihood of losing all offsite , power at nuclear plants or for similarly designed incoming transmission lines and switchyards at nonnuclear plants. In light of this limitation, the objective of this study was to derive some general frequency and duration characteristics which could be applied to nuclear power plant offsite power system designs and NUREG-1032 Appendix A A-22
site locations, generically or on a case-by-case basis, considering specific susceptibility to the v'arious weather hazards. The approach taken was to develop a range of loss of offsite power frequency and duration relationships based on weather' hazard rate and intensity and past operating experience. First, data for all loss of offsite power events in-volving both partial or total failures were reviewed. Included were weather-related total loss-of-offsite power events and significant partial loss-of-offsite power events, such as those causing the complete loss of power to or from a switchyard. These data are provided in Table A.7. Here again, as with grid reliability experience, this data base is rather small to derive plant location and design dependent conclusions regarding the expected frequency of loss of offsite power due to severe weather occurrences. Normally, regression analyses would be used to correlate failure rate, design factors, and weather hazards. However, the losses of offsite power are so rare that the available data are too limited to take such an approach. The method taken to correlate loss-of-offsite power frequency to weather hazards is a simplified approach similar to that taken to correlate transmission line and terminal outage data to various types of weather exposure (Reference 7). It has been assumed that the frequency of offsite power loss due to weather events is proportional to the weather hazard rates at a site. The weather hazard rate is a measure of frequency of conditions with the potential to cause loss of offsite power. The following weather hazard rate indicators were selected:
1. snow / ice - inches of snowfall, per year

2. tornado - frequency of tornadoes per square mile, per year

3. hurricane and wind - frequency of storms per year with wind speeds exceeding approximately 75 mph NUREG-1032 Appendix A A-23
s. I l l Table A.7. Severe Weather-Induced Losses of Offsite Power j Used in the Analysis , i I Total Losses of Offsite Power:
. . . . Site Date Duration (hours) Weather Type 1 ,
Point Beach 02/05/71 0.13 Snow / Ice - Fort St. Vrain 05/17/83 1.75 Snow / Ice Pilgrim 05/10/77 2.67 Snow / Ice
! Dresden 1 11/12/65 4.00 Tornado ,
Millstone 08/10/76 5.00 Hurricane i Pilgrim 02/06/78 8.90 Snow / Ice
~
Major Partial Losses of Offsite Power: l Site Date Weather Type ; i j Browns Ferry 03/01/80 Snow / Ice , D. C. Cook 02/04/78 Snow / Ice ! Pilgris 10/12/82 Hurricane i San Onofre 02/24/69 Hurricane Arkansas 02/22/75 Tornado l '
Nuclear One Arkansas 04/07/80 Tornado l ,
Nuclear One Browns Ferry 04/03/74 Tornado l l l 4 l i NUREG-1032 Appendix A A-24 [
Stores are classified as hurricanes when wind speeds reach 75 mph. The frequency of this wind speed was used as a correlation point to determine the variability of hurricanes and high wind hazards at various locations (sites). By dividing the number of losses of offsite power which have occurred by the , cumulative historical weather hazards for each weather type at nuclear power plant sites, an offsite power failure proportionality factor for each weather type was derived. This process can be represented as follows: Pg= 1 8 ji where P g is the pr'oportionality factor for weather type i Ng is the observed number of offsite power losses due to weather type i H yg is the cumulative weather hazard factor for weather type i at site j and Hyg = hjg at y where h), is the weather hazard rate for type i weather at site j and Atj is the cumulative site years since commercial operation at site J. i NUREG-1032 Appendix A A-25
The expectation frequency of loss of offsite power can then be computed by the following equation: Sji = Pyh jj i where S yg is the estimated frequency of loss of offsite power at site j for weather type i, and Py and h yg are defined as before.
! Based on the data from Table A.7 and cumulative weather hazards for U.S. nuclear plant sites through 1983, the following weather induced failure proportionality factors were estimated:
P 3jg = 1.8 x 10 / inches of snow fall P H/W
= 2.6 x 10-2/ incident P
T
= 27 square miles (for single right-of-way)
The weather hazard factors for each site were derived from National Weather Service data (References 8-13) where available. If data for a particular weather 4 type at a site were not available, the operating experience of that site was not included in the estimates. Normally this type of correlation activity would be supported by a statistical validity test. As pointed out previously, since there have only been a few teather-related losses of offsite power at nuclear plants, the statistical validity could not be ascertained. However, as a test of the reasonableness of . this formulation, a plot of cumulative weather hazard factor for each site (Hg ) versus total cumulative weather hazard factor tabulated for all applicable I nuclear plant sites (IH ) 4was made, and the severe weather-related operating cxperience for both total and major partial loss of offsite pcwer events were identified. A comparison was also made of the number of sites falling uithin subdivisions of the range of cumulative weather hazard factors. This information is provided in Figure A.6, where the number of 4 NUREG-1032 Appendix A A-26 I
tw mann 20 .
~ ,,10,000 47 1P r r :: - 15 . .
7.500 t I-
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ej 1P & s - Ei fj 5.. -- . 2,500 0 --- 0 10I 102 ic3 H S. . HURRICANE / WIN:r >75 mpn 20 . ' 20 1T. . 2P 15 - i- 15
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l TORf1ADOES 20, ,, 0.100 I - 4 IT . i 15, , 0.075 I
**
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Oi e, 10. - 0.050 . e i 6 . o . T - di!!
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losses of offsite power followed by a "T" represent total losses of offsite power and those followed by a "P" represent major partial losses of offsite power. Since frequency of loss of offsite power due to weather has been cssumed to be proportional to the magnitude of weather hazards, the occurrence of weather-related losses of offsite power should favor the sights with the highest cumulative weather hazard. In general it does. The events identified in Table A.7 are typified by durations of up to several hours. The failures are somewhat localized, isolatable, or repairable with modest effort. Design factors such as transmission line right-of-way sepa-ration, structural strength of transmission and switchyard components, insulation from effects of adverse environments, and operational factors related to repair capability or use of alternate, available power sources will
- impact the likelihood and duration of loss of offsite power events of this type. Events of the type just discussed will be referred to as severe weather events throughout this report.
None of the events identified in Table A.7 involved tornado or hurricane /high wind conditions which severely damaged structural elements of all transmission End/or switchyard components of sources of offsite power to the plant. Such an cccurrence, although rarely expected, could require many hours or days to repair and restore offsite power. The frequency of these more extreme weather-related power losses can be esti-mated by determining the frequency of weather conditions sufficient to damage all' offsite power sources. The same design factors noted above for the more repairable loss of offsite power events will determine the susceptibility, and e l l l i NUREG-1032 Appendix A A-29 i
l. . - , - - - - - . . - - _ - - . - _ . - - . . - - - - . . . - . . . - , - . - - _ . . _ - - . . _ . . _ _ . . . - . . - .
i l! l 1 thus frequency or hazard rate, of weather conditions which could result in l crea wide transmission and/or switchyard failures. Based on the National Electric Code, wind conditions on the order of 125 mph could cause extensive j power transmission losses although this will vary depending on design details. Tornado (es) must strike all rights-of-way or switchyards with sufficient - intensity to damage the minimum number of components required to supply offsite , power in order to cause a long duration loss of offsite power. The probability cf equipment failure given the occurrence of these extreme weather conditions is assumed to be unity, or nearly so, and thus the likelihood of offsite power - ! loss can be approximated by the frequency of occurrence of the extreme weather ! condition. The extreme hurricane (known as great hurricanes) and high wind 1 - frequencies are available from National Weather Service data. To estimate the frequency of single or multiple tornado-strikes damaging all. transmission lines or switchyards requires modeling of the offsite power transmission line l geometry (References 12 and 13) and using site / area data for tornado frequency, ! intensity and direction. This type of mechanistic, probabilistic analysts was not performed as part of this work. A simpler, bounding type of approach was used. The tornado-related loss-of-offsite power frequency for a single right- ! of way derived previously was used. For some sites the estimation of tornado ! caused loss-of-offsite power frequency could be overestimated by an order of l magnitude or more using this approach. When the tornado frequency is low, as ! l it is at most sites, it will not make a noticeable difference when computing - - l total loss-of-offsite power frequency. For sites in relatively high tornado i frequency locations, the results may be more appropriately treated as a high, j rather than a best estimate. For purposes of this work, the low estimate for , I tornado caused loss-of-offsite power frequency was taken as "negligable" l compared to the high estimate. This lower estimate would be indicative of ! sites with transmission line rights-of-way spreading out in directions obtuse I i each other. ! Events of the types discussed above are referred to as extreme weather events throughout this Appendix. Although occurrence frequency could be as high as 4 0.01 per site year, it will more typically be less than 0.001 per site year for ) extremely severe weather events. 1 . l NUREG-1032 Appendix A A-30
The time necessary to restore a source of offsite power for weather-related failures will be dependent on the severity of damage caused by the event. Major structural damage can typically require 8 to 24 hours or longer for repair. Data obtained from MAIN and MAPP (References 14 and 15) indicate that.it takes on' the order of 8 to 12 hours to restore transmission or terminal point outages due to severe weather. For this study, nuclear power plant outage time data for losses of offsite power due to severe weather were used to estimate restoration likelihood for the less than catastrophically damaging weather cvents. Data for total loss-of-offsite power events were fitted to a two-paraseter Weibull distribution and used to generate the restoration likelihood curve shown in Figure A.7. Also shown in Figure A.7 is an " enhanced" recovery curve which can be used to differentiate plants with practicable power restor-ation procedures for these weather types. The applicability of enhanced recovery would be dependent on the capability and procedures to restore power within about 2 hours for a given weather hazard. An estimate of the total severe weather-related frequency of loss of offsite power was derived by summing the values for each weather hazard type at all nuclear plant sites. Plant-specific design or procedural details can affect the weather-related loss of offsite, power frequency estimations. Therefore, an approach was taken to derive the range of possibilities rather than attempt to provide site-specific estimates. It should be noted, however, that not all weather hazards could be accounted for at every site due to lack of data. More-cver, some weather data extrapolations were necessary when weather station data were not nearby a site. The frequency range derived was large, and to determine where a particular site / design combination would fall in that range requires cvaluation of the site-specific details identified previously. For the purpose of this work, the range was subdivided into groups with approximately a factor of three difference in median frequency. The subranges so derived are provided in Table A.8. This partitioning was made to allow generic evaluation of severe weather hazard effects on loss of offsite power frequency while at the same time providing perspective on the potential for plant-specific differences. Figure A.8 shows the severe weather frequency and duration combinations corresponding to the groups defined in Table A.8. NUREG-1032 Appendix A A-31
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Table A.8. Severe Weather-Induced Coss of Offsite Power Frequency / Recovery Severe dectber.-Induced Loss of Offsite Power Frecuencv(S) Frequency of Severe Weather-Induced Frequency Group (5) ' Loss of Offsite Power 51 Less than 1 per 350 site-years (o.coe/ye) 52 > 1 per 350 site-years and 71 per 120 site-years (o.. .s /yr ) 53 Greater than or ecual to 1 per 120
. site-years ( o. o s a fy r) . Recovery (R) .
Recovery from Sever Weather-Induced Less of Offsite Pcwer Grcups (R) Recovery , Capability R1 Plant has capability and procedures to recover offsite (non-emergency) AC
. power to the site within 2 hours following a severe weather-induced Icss of offsite power.
R2 All other plants not in RI. Severe Weather-Induced Loss of Offsite Pcwer Frecuency/Recoverv ('SR) Severe Weather-Induced Loss of Offsite Power Frequency / Recovery Group (SR) Frequency Group (S) RecoveryGroup(R) ! SR1 51 R1 SR2 52 R1 SR3 53 R1 l SR4 51 R2 l SR5 ' 52 R2 i SR6 . 53 R2 l NUREG-1032 Appendix A A-33 -
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Fcr extremely severe weather caused loss of offsite power due to great hurri-canes, very high winds (greater than'125 mph), and major damage from tornados, restoration of offsite power was not assumed to occur within ;4 hours. The frequency breakdowns derived in a manner similar to that for severe weather are provided in Table A.9. Again it must be noted that a site-specific assessment of the susceptibility to these weather hazards must be performed to determine l the site-specific expectation frequency. Generic Loss of Offsite Power Correlations Combinations of design, grid, and weather factors derived in the previous sections provides a wide spectrum of possibilities for loss-of-offsite power frequency and duration. Each of these factors was subdivided to account for known or hypothetical but reasonable differences in frequency and duration; typically, a factor of two to five difference was maintained for these sub-divisions. The intent was to develop a discrete set of frequency and duration groups which could account for actual and potential differences in both design cnd location (grid and weather) for the spectrum of nuclear power plant sites. The frequency of losses of offsite power lasting duration "t" or longer can be
,cstimated by appropriate combination of the correlations which were developed J
in this Appendix and can be represented by the following equation: A LOP It)
I (t) j + SR k(t) + SS) i + GR (t) where Ij (t) is the plant-centered loss of offsite power frequency correlation defined in Table A.3 and Figure A.2; GR y (t) is the grid-related loss-of-offsite power frequency correlation defined in Table A.6 and Figure A.5; SRk (t) is the severe weather-related loss of offsite power frequency correlation j
defined in Table A.8 and Figure A.8; and SS) is the extemely severe weather-related loss-of-offsite power frequency , define in Table A.9. NUREG-1032 Appendix A A-35
Table A.9. Extremely Severe Weather-Induced Loss of Offsite Power Frequency Extremely Severe Weather-Induced Less c' 0'f site Power Frequency (55) .. [ Frecuency of Extremely Severe Weather-
Frequency Groups (SS) Induced Less of Offsite Power
's-sa-SSI Less than 1 per 3500 site-years (0,ooop)
SS2 > 1 per 350C site-years and 71 per 1200 ' site-years (0.WF N f SS3 > 1 per 1200 site. years and 7 *. per 350 site years (o.coJ/gN SS4 > 1 per 350 site-years and
< 1 per 120 site-years (04/fe3 $55 Greater than or ecual to 1 per 120 site-years (o,ots/y) ,
i i l - l NUREG-1032 Appendix A A-36
= - - _ - .
The identif,ication of the I4 factor is the most straightforward since it is based on configuration. As a first cut, the appropriate GR) factor can be identified by dividing US nuclear sites into two categories: (1) FPL sites, GR7, and (2) all other sites representing average frequency expection of grid failure, GR1'or GR2. The SR k and SS) factors are not so easily identified since both design specifics and hazard rate must be determined. It is possible, however, to bracket these factors with a range that can be used to judge importance of station blackout considerations using hazard rates and proportionality factors for severe weather, and using the more extreme hazard rate for extreme weather. A test of the loss-of-offsite power correlations which were developed was made by comparison with plant-specific results from published PRAs. Figures A.9 through A.13 provide these comparisons. With th3 exception of the Zion PRA, the results show reasonable agreement. The cross-hatched area represents the high and low estimate for extreme weather-related losses of offsite power. The differences with the Zion PRA results could be due to one of several possibilities: design and prgcedural factors which are more reliable than assumed in the comparison; the Zion PRA results are optimistic; cr the models and correlations derived for generic analyses have limitations when applied to some plant-specific cases. It is because of these considerations that a generic analysis must be used with caution in plant-specific applications. However, the generic models can usually provide good
" ball park" results for generic applications and perspectives. Clearly the more details available and included in the models regarding design, procedures, alternate power sources and protection provided from severe weather conditions, the more likely that the generic results will closely equate to plant-specific results.
NUREG-1032 Appendix A A-37
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NUREG-1032 Appendix A A-42
The development of a more limited number of generic loss-of-offsite power frequency and duration relationships, which could be used for regulatory analysis, involved the clustering of the site / design factors to determine if combinations of these factors could be grouped into a more limited, but still representative, set. A set of nine cluster groups was derived from the set of site / design possibilities using the Fastelus procedure of the SAS package (Ref. 4). In order to limit the number of cluster groups, the clustering had to be based on loss of offsite power durations of 4 to 16 hours. Figure A.14 provides a plot of the cluster groups derived from this analysis, and Table A.10 identifies the factors that can be in each cluster group. Because design, grid, and weather all play a role in the frequency and duration relationship for each cluster, it is difficult to generalize about the dominant factors affecting loss of offsite power. It is possible to say that the higher frequency at longer duration groups (clusters) are most heavily influenced by weather hazard susceptibility. It is also reasonable to speculate that perhaps no known nuclear plant has the combination of site, location and design features which would result in the highest frequency and duration correlation developed in this study (e.g., Cluster 9). ! l l NUREG-1032 Appendix A A-43 i 1
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ML20136G019

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