ML20149F812

Forwards Draft Commission Info Paper Re Source Term, Incorporating Comments Received to Date for Concurrence or Comment by 851031

ML20149F812
Person / Time
Issue date:	10/24/1985
From:	Bernero R Office of Nuclear Reactor Regulation
To:	Eisenhut D, Speis T, Thompson H NRC
Shared Package
ML20149B718	List: ML20149F619 ML20149F626 ML20149F650 ML20149F663 ML20149F678 ML20149F687 ML20149F777 ML20149F792 ML20149F799 ML20149F812 ML20149F826 ML20149F830
References
FOIA-87-743 NUDOCS 8802170373
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Text

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Mo og 0,, UNITED STATES

.t ,8 a NUCLEAR REGULATORY COMMISSION

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OCT 2 41985 i

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'$ , MEMORANDUM FOR: Distribution List 9

g ,. FROM: Robert M. Bernero, Director 4

Divisicn of Systems Integration

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SUBJECT:

CONCURRENCE IN SOURCE TERM COMMISSION INFORMATION PAPER

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1. On September 27 Daniel Muller furnished you a draft copy of the subject paper for your information and comment. The attached concurrence copy, which has incorporated connents received to date, is provided for your parallel review prior to being routed for Office Director concurrence, si Please concur, or furnish coments, to me (ext. 27373) or Jerry Hulman w'

-" (ext. 27763) by COB October 31 to ensure the WITS (Item 840401) schedule *

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is met.

v _c obert H. Bernero Director Division of Systems Integration

Enclosure:

As stated cc: D. Ross, RES E. Jordan, IE

0. Bassett, RES F. Gillespie, RES Z. Rosztoczy D. Matthews, IE H. Silberberg, RES '

Distribution List Darrell Eisenhut

> Themis Speis Hugh Thompson I

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• NRC Form 8 U.S. NUCLEAR REQtJLATORY C08iendl3SION

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NRcW o24o RECOMMENDED DISTRIBUTION

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~ Staff Plans for the Use of New Source Term Information in the Licensino Process

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.' 00T 3 41M LGHulman JMitchell MEMORANDUM FOR: Distribution List g. o o W Robert M. Bernero, Director LSoffer JMalaro

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Division of Systems Integration TQuay MSpangler

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SUBJECT:

CONCURRENCE IN SOURCE TERM COMMISSION INFORMATION PAPER

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Ni On September 27 Daniel Muller furnished you a draft copy of the subject 7;( paper for your information and comment. The attached concurrence copy,

?/'. which has incorporated comments received to date, is provided for your Je parallel review prior to being routed for Office Director concurrence.

Please concur, or furnish connents, to me (ext. 27373) or Jerry Hulman N. . *

(ext. 27763) by COB October 31 to ensure the WITS (Item 840401) schedule N.; <

is met.

Origiul 5132ed 375,

'.',' F.obert M. Bernero.

.. Robert M. Bernero, Director

" - Division of Systems Integration

Enclosure:

As stated cc: D. Ross, RES E. Jordan, IE

. O. Bassett, RES F. Gille'spie, RES Z. Rosztoczy D. Matthews, IE M. Silberberg, RES Distributich List Darrell Eisenhut Themis Speis

- Hugh Thompson a

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DATE : 10/ d /85- : 10/t.'/85 :.10/l(/85 :  :  :  :

OFFICIAL RECORD COPY-

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SE From: William J. Dircks ,

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Subject:

STAFF PLANS FOR THE USE OF NEW SOURCE TERM INFORMATION IN THE p.7 LICENSING PROCESS tp F. .'

Purpose:

To inform the Comission of current and forthcoming staff 6 activities for use of new source tenn information in the regu-

'9 '

latory process, to inform the Comission of preparations for l'T Comission actions based upon new source term information, and to solicit Comission's coments on the direction and priorities

, j of these staff a.ctivities.

. Sumary: A recent staff document NUREG-0956, Reassessment of the Technical Basis For Estimating Scurce Terms, sumarizes and assesses -

recent advances in source term research; that is, prediction of the quantities of radioactive materials transported and released in reactor accidents. The ability to predict, or to ass 0re conservative assessments of these releases, lies at the very heart of the licensing process. We measure the need for and the extent of protection of the public from undue risk by careful assessment of possible accidents and the releases of radioactive materials they might cause.

Accident release estimates have been used in the regulatory process for more than two decades. Many parts of the process o, , are based upon release assumptions contained in the 1962 document "Calculation of Distence Factors for Power and Test Reactor Sites" (TID-14844), which forms the basis for 10CFR100, as well as those based upon the more recent (1975) "Reactor Safety Study" (WASH-1400)

J risk estimates. Examples of regulatory uses of TID-14844 include containment performance, environmental qualification of equipment, air filtration and other fission product mitigation systems, accident monitoring onsite and offsite, as well as siting.

Examples of regulatory uses of WASH-1400 release estimates include emergency planning, advising Congress on Price-Anderson

,c Act insurance, evaluating offsite impacts and risks (including

, Environmental Impact Statements), assessing offsite contamina-tion and recovery, and investigating new regulatory requirements.

1

Contact:

R. M. Bernero, NRR, 492-7373 or L. G. Hulman, NRR, 492-7763 ,

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O Although revised so'urce terms valid for all reactor types hi' and accident sequences cannot be defined at this time, the '

M accumulation of substantial new infonnation on the subject N compels us to re-evaluate those regulations, standards .

((- and practices which are based on such estimates.

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'i The regulatory staff has followed the research on this subject

?? very closely, particularly on watch for evidence which might f '.i indicate gaps or weaknesses of safety significance in our present

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. regulations. As the staff informed the Conraission on March 14, 1985, we have not found any significant deficiencies in the

<9 r regulations based upon our knowledge of this new infomation. We

l- are using the new infomation in the search for risk outliers, P cost-benefit assessments, standard plant reviews and in environ-

!- mental impact statements. We have also tracked the research to

• .i identify areas in which adjustments ir, regulations, standards .

and practice may be desirable. We have found several such areas y and have summarized them below.

V Discussion: 1. Regulatory Uses of Source Term Information - Accident source tenn infomation is used in a nunder of regulatory areas. The 10 areas for which the staff believes adjustments are merited based upon the emerging research, are discussed in the enclosures and are suaynarized as follows:

a) To search for any severe accident scenarios which pese disproportionate risk. Under the Severe Accident Policy Statement, industry and the staff will search for risk outliers. This activity is to be accomplished through the

'. proposed Industry Degraded Core Rulemaking (IDCOR) Group reviews. The uses of new source tem information in calibrating risk levels for several reference plants, and in identifying any risk outliers at every plant, are to be sunnarized in a separate Connission Paper.-

.. b) To assure acceptable leakage of accidental releases from containments. The leakage criteria are presently based

.: upon the use of TID-14844 source tem assumptions, and are implemented through 10CFR50, the Standard Review Plan (SRP),

and related Regulatory Guides. Need far modifications involv-ing both potential decreases and increases in regulatory oversight have been identified. The discussion of this .

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area, including staff plans for evaluating a potential decrease in regulatory criteria for containment leak rate i ,' . testing, and an increase in regulatory oversight for N increasing assurance of containment, integrity, is presented Odf; in Enclosure 1 as Item 2.

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c) To confim the adequacy of the environmental qualification of electrical equipment. Present radiological qualification

[pJ. is based upon the use of TID-14844 release assumptions, as

'. . articulated in related Regulatory Guides and the SRP.

,9 Present practice in this area and staff plans for further ii,.d"' study of the adequacy of present qualifications for the range of conditions indicated by the present and forth-l', . coming research are~ discussed in Enclosure 1, Item 3. .

, d) To improve emergency planning. This area is presently regulated through 10CFR50.47, Appendix E to 10CFR50, and J.

an interagency agreement with FEMA. Bases for regulation

.-.'^ include the range of accidents that could occur as assessed primarily using WASH-1400 source term estimates. '

c. The staff proposes to reevaluate the bases for emergency planning guidance, to coordinate with FEMA and affected state i

, and local entities, and potentially to propose modification

. tc emergency planning requirements in 10CFR50. Staff plans in this area are summarized in Enclosure 1, Item 4. '

e) To assess potential accident impacts to judge whether onsite 4

property insurance is adequate, to advise Congress on the

, adequacy and need for continuation of offsite indemnification requirements (Price-Anderson Act), and to evaluate environ-mental impacts under the National Environmental Policy Act (NEPA). Three activities are proposed; 1) detemining the

. adequacy of onsite insurance requirements as implemented through 10CFR50.54(w); 2) detemining the influence of the new research on the reconsnendation to Congress that the 1

possibility of low likelihood, high consequence accidents c' warrants an increase in the offsite liability limits; and

3) using the new source tem information in environmental assessments of accidents for licensing evaluations.

Regulatory guidance is given in 10CFR140, 10CFR50,

- 10CFR51, and a 1980 Conunission Interim Policy Statement relating to accident considerations under NEPA. This area is discussed in Enclosure 1 Item S.

f) To confim that accidental releases of fission products will be adequately mitigated by the use of engineered safety features.

The criteria in this area are presently based on the use of-n TID-14844 release assumptions, and are implemented through 10CFR100 guidelines, 10CFR50 design criteria, related Regula-f_s tory Guides and the SRP. The criteria include provisions for protection of the publ.ic and reactor operators (10CFR50,

. Appendix A, General Design Criteria 19). A discussion of 2

present practice, and staff plans for considering potential -

short tem and subsequent changes in related regulatory ar- f p < n. .

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N 4 CD guioance, are presented in Enclosuro 1, Item 6. To prevent licensees from incorrectly expending valuable resources in  :

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gr implementing approved backfits related to control room

. habitability involving radiciodine that could more productively be u~ sed in other' safety activities, a draft Ger.eric Letter 4'? is being prepared explaining a new staff view. Another short

-I' term use of the research is a potential relaxation .of guidance T.i? for injection of caustic spray additives on PWRs for which a V draft Branch Technical Position (BTP) is being prepared.

Both of these short term actions are being processed

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by the staff for transmittal to licensees and applicants.

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g) To assure that adequate onsite and offsite instrumentation is 1(.;,

available to aid in preventing or mitigating releases, and 2, for evaluating the consequences of accidents. This area is

, regulated through 10CFR50, related Regulatory Guides, and f the SRP. Discussion of this area is presented in Enclosure 9: 1 Item 7. ,

i h) To assure that adequate provisions are available for coping r with offsite post-accident contamination and recovery.

Guidance in this area flows from 10CFR50, the views &nd '

practices of other Federal agencies and general practice.

Staff plans for use of the new source tem information in

,. policy considerations related to the use of. thyroid blocking '

agents, and in standard plant reviews, are discussed in Enclosure 1, Item 8.

1) To perfom safety evaluations, to evaluate the safety e- significance of generic issues, and to judge whether back-fits are necessary. First, present safety evaluations are .

l prepared by evaluating site characteristics, plant features

% and operating procedures for a set of postulated design b basis accidents. The worst fission product release assump-L' tions used in the suite of design basis accident analyses are those associated with TID-14844. Alternatives to the present  !

( methodology are proposed for study. Second, the present l- methodology for judging the safety significance of proposed ,

generic issues is based on WASH-1400 source term estimates. '

.. The overall methodology allows a rough but adequate screening i of the issues. Reprioritization of existing generic issues e

a using other source tem estimates would not serve to provide a fU significantly improved screen becaum of the relative i.

coarseness of prioritization proce:s, and is not planned.

,1 Lastly, evaluations of the need for backfitting can be improved 7; , through the use of the new research results and are planned.

Lc These three subjects are discussed in Enclosure 1 under Item 9. ,

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', j) To assess the adequacy of site and plant characteristics.

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". .-j Present criteria flow from 10CFR100,10CFR50, various

@,% 3 Regulatory Guides and the SRP that couple both sets of

(,[ characteristics. Rulemaking on siting issues was begun in

't: 1980, but was held in abeyance pending the availability of Nt source term research results. A reconnendation for i; uncoupling plant and site characteristics and reinitiating rulemaking is presented in Enclosure 1. Item 10.

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.? Each area is discussed in detail in Enclosure 1. The staff

.5, schedule for implementation and a preliminary programatic f7.f.

cost-benefit are presented in Enclosure 2.

2. Severe Accident ~ disk Decisions - Prior to the THI a'ccident

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in March,1979, considerations of accidents more severe than

.7, those used in reactor licensing reviews were limited to generic fN assessments. The primary generic evaluations were those associ-5' ated with WASH-740 in 1957, those debated in developing 10CFR100 published in 1962, and those associated with L. WASH-1400 published in'1975. Following the TMI accident, an '

action plan was developed to implement the lessons learned from the event. -Consioeration of the evaluation of accidents more severe than those used in licensing evaluations (called severe accidents or Class 9 accidents) were included in the action plan. The October 2,1980 Advanced Notice of Proposed

.. Rulemaking on Consideration of Degr'aded or Melted Cores in Safety Regulation (45FR65474), the hydrogen rules, the severe accident policy (SAP)~ devcinpment, the Interim Policy State-ment on Severe Accident Considerations Under NEPA, control roomoperatorprotectionfromsevereaccidents(presentlya generic issue awaiting decisions on source terms) and the presently inactive July 29, 1980 Advanced Notice of Ru'emaking

. related to siting (45FR50350) all followed. The essential basis of the proposed SAP is that, except for possible risk outliers and refinements in regulatory oversight, existing plants pose no undue risk to the public. Furthermore, it was concluded that future reactors can be refinements of existing designs with low risks, and the search for risk outliers at existing reactors can a be done in an orderly manner. The staff has concluded that the

, proposed IDCOR evaluations of existing reactors, and the staff review thereof, can be the orderly search for outliers. This

. process is to be discussed in a forthcoming Commission Paper. '

/. 3. Activities and Interactions - The proposals outlined in the

, enclosures constitute the staff's current plans for use of

.' completed and forthcoming source tenn research in the regulatory jn process. A fundamental assumption used in the preparation of

, this agenda is that interactions with the public and industry

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ed including the remaining confirmatory activities, and its

, i' application. -The interactions will include publication of results in NUREG documents, requests for and resolution of 8: '

. coments from the public and industry on proposed actions, and am)1e peer review to ensure adequate technical bases for decision 61 macing. Internal NRC reviews by NRR, IE, RES, ACRS and CRGR y.j, will also be scheduled, e.-

4' A significant amount of dependence exists between the risk

.:l-rebaselining activities by RES and IDCOR, and the search for N- plant specific severe accident risk outliers. Furthermore, many 7.!

• remaining experimental, code development, code verification and c' uncertainty research activities are directly related to the V . ability of NRR and IE to implement research findings. I conclude he that the fr'amework for the orderly implementation of source term lf;- and related research into the regulatory process described herein f provides for the necessary timely interactions between the program offices, and between the staff and the Comission.

. , Finally, the staff offers the agenda in this paper for whatever coment or direction the Comission may wish to offer. ,

Scheduling: This paper may be presented to the Comission in open session.

William J. Dircks Executive Director for Operations

Enclosures:

1. Regulatory Areas and Potential Changes

2. Preliminary Cost / Benefit and Schedule for Change 9

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1 ENCLOSURE 1 l m

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d REGULATORY AREAS AND POTENTIAL CHANGES .

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& A review of existing areas of regulatory practice influenced by accident .

9 source terms was used to identify areas where the* staff concludes changes

,?.i may be appropriate. Ten areas have been identifiedi For each area 5 identified, a description of the background and current practice, .

@. possibility for and character of potential changes in practice, costs,

benefits and schedule follows

1. IDCOR-NRC STAFF SEARCH FOR RISK OUTLIERS k!2 Background and Current Practice: Under the Comission's Severe Accident Policy

7.

• 5tatement, a search for plant design features which might contribute inordi-s natelytorisk(riskoutliers)atexistingreactorsistobeundertaken.

.: This search is to be accomplished by the development and application of a  !

methodology for an industry-wide plant-by-plant search. The methodology 3 involves developing techniques for evaluating risks using the source term and .

'; severe accident research at a small number of power reactors used as reference i plants, and extending the lessons learned to all existing power reactors.

Industry has been active through the Industry Degraded Core Rulemaking (IDCOR) 1 3 group, and the staff has, followed and commented on their activities. The staff has also developed a plan for related research and application activities wiiich l'

are to be reported in a separate Comission. Paper..

Possibility for and Character of Changes: Source tem and related severe i

,- accident research is to be used in rebaselining the risks at a small number of t reference plants. These plants represent the spectrum of LWR designs used in

- the U.S. Industry is to proceed to search for risk outliers at all reactors.

7

,3 The use of the source tem research results are expected to result in generally  !

t equal or . lower estimates of risk than with the use of the WASH-1400  !

N methodology.  !

l' U '1 Costs and Benefits of Chance: Intangible benefits arise from the use of i P contemporary source term related research that fully considers the chemical and

._ physical processes involved in the release and attenuation of fission products L in core melt accidents. The result should include more accurate identification i I: of risk and any backfit requirements therefrom.  :

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] Tentative Agenda and Schedule: To be discussed in a separate Comission Paper. ,

j,, 2. CONTAINMENT PERFORMANCE -

Background and Current Practice: The current Technical Specificatio.1 leak H., rates of containments ran Licensees are l required by 10CFR50.54(0)ge from an to perfom 0.1% to 2.0%leak integrated by volume rate testperin day.

accordance  ;

4' with 10CFR50, Appendix J. These tests are perfomed at least once every forty '

months to demonstrate that the actual leak rate is less than 75% of the design su*

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2 ENCLOSURE 1 b.d .

5 basis value contained in the technical specifications of each plant's license.

.$ In addition, the gas and liquid leakage from equipment and components that are -

b required to handle fluids and gases containing radioactive materials Q accidentoutsideacontainmentisaddressed(TMIActionItemIII.D.1) post-by a leak p detection program. Furthermore, criteria used by the staff to judge licensee d venting and purging operations while at power also consider accidents '

pl involving the use of TID-14844 fission product release assumptions.

Y.' The Technical Specification leak rate of a containment is established by Y. calculations which assume that the post accident airborne concentrations of

'O iodine vapor and noble gases specified by TID-14844 are instantaneously '

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• released and unifonnly distributed in the containment atmosphere. These  :

4 calculations deliberately ignore, or treat simplistically, the removal of '

2e iodine by such natural processes as dissolution in water, chemical reactions, t

.l and plate-out. They also neglect release of other fission products such as

. cesium, tellurium and strontium, and non-radioactive aerosols which could be

[; in the containment, or any undetected breach of containment.

In progress at this writing are proposed changes to Appendix J that do not '

incorporate new source term information. The Federal Register notice for these changes, however, encourages the submission of source term related connents for  !

( a future change in Appendix J.

N Possibility for and Character of Changes: The assumed release of iodine vapor during reactor accidents currently dominates the calculations of offsite doses

• from which technical specifications of containment leak rates are derived.

Current research indicates that most iodine released in core melt accidents  ;

will be in chemical and physical forms other than iodine vapor. Further, such l

releases will include significant fractions of such isotopes as cesium, tellurium, and strontium in addition to iodine. Thus, the present staff  ;

assumption that elemental iodine dominates offsite risk is not correct. It is  ;

also now known that aerosols may be released following core melt accidents, i and that containment by-pass or failure is of relatively greater importance to public risk than diffusive leakage. Thus, the basis for specifying containment ,

e '. leak rates should be re-avaluated, and a greater emphasis should '

,' '. be placed on maintaining the integrity of containments.

L In the near-tenn, the staff intends to study the advantages and disadvantages

t. of alternative containment leak rate test requirements and acceptance criteria  :

in terms of accident risk significance as well as operational considerations. l' Early indications are that these leak rates might be relaxed by factors ranging from two to ten, depending upon the present value, without signifi- ,

cantly impacting public health and safety. Also to be considered are accident management techniques that would delay or prevent containment failure during i:'

,f severe accidents. j On the basis of this :,tudy, the staff may propose revisions in the Standard ReviewPlan(SRP)criteriaandtheAppendixJtestrequirementsandacceptance b([

l criteria. Candidate revisions include:

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3 ENCLOSURE 1 f .

.,' a. Specifying a containment leak rate value and testing frequency for each of ,

L.? the major types of containment. ,

o b. Adopting administrative controls or penetration design guidelines to aid i gg in precluding 'an undetected breach of containment integrity. ,

3, d c. Requiring testing to_ provide additional assurance that an undetected

. breach of containment integrity does not exist (inherent design / operation

'%a features such as the need to maintain a subatmospheric condition or inerted atmosphere may be readily adapted for this purpose).

L* d. Modifying staff criteria for valve closure times in lines which penetrate m' ,

containment and may be open during operation.

b . To assess the extent of any proposed revisions, the following studies will be undertaken:

a. Perform mechanistic analyses (releases from fuel, plate-out.. spray i washout, attenuation along release path, reemission etc.) of fission product releases from the core and transport within the containment.

b. Evaluate threshold levels of containment leakage important in risk estimates. .

c. Explore the viability of implementing testing practices that provide a continuous indication of containment integrity.

d. Provide bases for adjusting Limiting Conditions for Operation and Action Statements in the Technical Specifications (i.e., to reflect their importancetosafety).  ;

For the longer term, the staff proposes to:

a. Explore the use of containment venting schemes for PWR's (similar to that proposed for BWR's) to reduce public risk by preserving containment integrity,

b. Examine the desirability of certain existing surveillance / maintenance practices (such as performing local leak rate tests on containment isolation valves during operation modes 1 to 4) that might lead to inadvertent ment sprays)and ,

post accident develop safety system an operability check lockouts method for(such as the contain-preventing inadvertent safety system lockouts.

Costs and Benefits of Change: Exploring the risk significance of alternative leak rate criteria in light of recent source tem research is expected to

+ require moderate staff resources in the near term. Longer term efforts, especially those for investigating approaches to enhanced containment integrity, may require a greater involvement of staff resources.

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j- 4 ENCLOSURE 1 P Increases in allowable containment leak rate of the order of factors of two to M ten are expected to offer direct and significant regulatory relief to many J operating licensees, since compliance with the criteria of Appendix J is costly y -

and is incurred regularly. The expense involved in the testing becomes larger

,, as the leak rate to be demonstrated becomes smaller. Furthermore, changes in

.- monitoring criteria can be expected to reduce the frequency of integrated W, tests.

Imposition of additional containment integrity requirements would have costs

." ranging from relatively low for routine administrative controls to high for

( systems which continuously indicate the status of containment integrity. The benefits would be primarily those associated with reduced risks, including .

improved accident management capabilities.

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" Tenative Agenda and Schedule: The staff's near-term study and reconnendations are expected to begin in FY86 and to be completed in FY87. Longer term studies C will commence in FY86 and are expected to be completed in FY88 or FY89. .

3. ENVIRONMENTAL QUALIFICATION OF EQUIPMENT Background and Current Practice: 10CFR50.49 requires electrical equipment important to safety to be capable of remaining functional during and following design basis accidents for witch that equipment is needed. The radiation environment assessed for this equipment is derived from the TID-14844 assump- ,

tions, while the temperature and pressure environment are derived from other .

accident analyses not involving substantial fuel melting. In addition to the source term research, there is a great wealth of information emerging from the examination of equipment subjected to the accident environment at Three Mile Island Unit 2 that can be useful in assessing operability.

Possibilities for and Character of Change: The radiation dose rates and integrated dose values arising from new source tenn research are likely to differ from those derived from TID-14844. The new infomation may show that equipment would be exposed to lesser quantities of iodine than previously postulated, but exposed to other fission products presently neglected. At the present time it is not clear whether the source term research will lead to an increase or a decrease in the accident radiation, temper ture, pressure, moisture and particulate environment for equipment qualification purposes. The staff estimates, however, that the present degree of equipment qualification,

derived from TID-14844, provides a substantial level of protection for many I severe accident conditions. A study will be perfonned to evaluate the I

differences between the existing equipment qualification criteria compared with those derived from new source tenn information. The infonnation emerging from i 4 examination of the Three Mile Island Unit 2 equipment will also be considered.

If the study indicates that the new source term information would result in a higher level of radiation for some accident sequences, then the study will also address the risk significance of having equipment qualified only to the lower level.

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Costs and Benefits of Changes: The staff estimates a moderate to high effort

~M will be required to evaluate the differences compared with the present criteria, and to address their risk significance with respect to the new source W

tenn research.

F. If the present criteria prove to be conservative, the staff anticipates no

ls direct relief to the industry, since there would be no great impetus to replace presently qualified equipment. However, at future maintenance outages, s equipment capable of being qualified to less harsh radiation environments, could be used to replace present equipment. Also, in some cases, it may be possible to extend the qualified lifetime of equipment. It is unlikely that qualification to moderately lower radiation levels would significantly reduce

', costs.

If the present criteria are non-conservative in radiation level, but have little risk significance associated with earlier equipment failure, the staff does not anticipate requiring licensees to take any innediate corrective action, such as replacement of equipment or its radiation sensitive components.

The staff, however, might require equipment or components of equipment qualified to higher radiation levels to be installed gradually as older equipment or components are replaced. This could potentially represent a large cost to the industry, depending on how non-conservative current' criteria may be, the number and type of components involved, and how much further testing would be re, quired.

. If the risks from exceeding the present criteria (radiation level', temperature, pressure and aerosol loading) are judged significant, the expected impact on the industry would be high since prompt corrective action and/or shutdown might be required. The impact of any increase in radiation levels above current criteria would have to be assessed on an equipment specific basis; i.e.,

different items of equipment are currently shown to >e qualified for different levels of radiation.

Tentative Agenda and Schedule: A staff study to assess the existing equipment qualification criteria will be initiated sometime in FY86 and completed in FY88. Although the staff can assess the change in criteria on a generic basis, involvement of industry may be required to assess the impact of a criteria change on specific items of equipment.

It should be noted that changes in magnitude and duration of the temperatures and pressures associated with severe accidents could have a substantial impact on qualification.

4. EMERGENCY PLANNING Background and Current Practice: Current regulatory practice in emergency planning is based upon consideration of a spectrum of accidents ranging from relatively frequent but low consequence events (including design basis accidents), to low probability severe accidents. The basic document upon which

6 ENCLOSURE 1 i

f e- the planning recuirements are based, NUREG-0396, examined severe accident consequences anc probabilities using WASH-1400 source' terms. Subsequently, a

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regulation setting forth a generic approach to onsite and offsite planning and

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preparedness (10CFR50.47andAppendixEthereto)waspromulgated. This regulation established generic requirements for a plume exposure pathway

'. - emergency planning zone (EPZ) of about 10 miles in radius, and an ingestion o.J pathway EPZ of about 50 miles. The sizes of these zones were considered 3

sufficient for the planning of various possible protective actions at any given nuclear power plant. A companion guidance criteria document (NUREG-0654),

prepared jointly by the NRC and FEMA, promulgated guidelines that included both s- evacuation and sheltering as potential offsite emergency responses. Since the NRC emergency planning requirements include actions by state and local govern-ments, the NRC depends upon FEMA to assist in providing an evaluation of the

. state and local offsite plans and preparedness around each reactor site.

Possibility for and Character of Change: As a result of the generic approach, misconceptions have arisen that all persons living within the plume exposure EPZ are at high risk while those living beyond 10 miles are safe from radiatjon exposure. In addition, some have interpreted the regulation as virtually always requiring evacuation to 10 miles or even further. The application of WASH-1400 source terms to study emergency planning has indicated that,the .

potential for early injuries and fatalities from severe accidents shows a large variation within the EPZ; that is, the risks are much greater in the inner portion close to the reactor than at the perimeter of the 10-mile EPZ. The concept of a graded res~ p onse that would recognize such a risk variation with distance and time was beginning to gain support at about the time the peer review process for the revised source term research began. Information obtained from the new source term research clearly supports and confirms this concept. Furthermore, the results of the source term work indicate that there may be a reduction in offsite accident impacts that may be specific to plant and site characteristics. Thus, employing ' generalizations or generic require-ments for all plants may not be technically valid. The source term research also indicates that severe accident releases may be more time dependent than those calculated using the WASH-1400 methodology, a conclusion which could affect the timing of any needed emergency actions. This discussion suggests that the current emergency planning regulation (10CFR50.47) should be re-evaluated.

Proposed Actions: A sequence of activities is proposed to implement changes in requirements and practice as follows:

a. Coordinate (IE)withFEMAandsolicitassistanceinmeetingswith appropriate state and local officials to explain the "graded response" approach to emergency planning and the possibility of classifying plants by groups by related accident consequence and risk estimates.

b. Provide (RES) an updated risk profile, including results of new source

' term research, for each of six (6) reference plants representing the spectrum of LWR designs used in the U.S.

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i 7 ENCLOSURE 1

.. c. With FEMA's assistance, determine (IE) the impact of a proposed change to '

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- emergency planning as discussed above, including appropriate protective l f,, , action response strategies. 1

d. Draft (RES and IE) revisions to 10CFR50.47 and Appendix E, taking into M consideration the revised source tenn methodology, possibly using the following alternative approaches

(1)classifyplantsorgroupsof f.; plants according to their apparent risk profiles, (2) use site specific ,

,. , ; emergency plans, or (3) use the graded response concept, s .s, .

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e. (IE,RESandFEMA)usingaconstituencydevelopedonanalt'ernative,begin .

1 a parallel effort to rulem king to revise guidance contained in the joint

FEMA /NRC NUREG-0654 document.  :

e:  :

f. Provide (IE, RES and NRR) detailed technical analyses and rationale for

. change as a result of the new source term information.

g. Coordinate (IE) with EPA to ensure that NRC actions are coordinated with the EPA ongoing effort to revise the Protective Action Guides.

Costs and Benefits of Change: Changes to emergency plannin'g are expected to require moderate to large staff resources. Significant efforts will be required to develop the technical bases for change resulting from revised source terms and risk profiles, and to coordinate with other federal agencies,

. ,such as FEMA and EPA, and with affected state and local agencies. Rulemaking activities, specifically with an anticipated hearing, are expected to require a .

moderate expenditure of staff resources.

Tangible benefits in terms of direct regulatory relief to most operating plant licensees are expected to be low, since the staff anticipates no major changes in licensees' emergency response capabilities or facilities. Intangible benefits in terms of reduced risk percsptions are anticipated to be high, and I tangible benefits may be realized for some licensees when it can be. established ,

that smaller plume and ingestion EPZ definitions may be warranted. Benefits to state and local agencies involved in emergency planning and response are expected to be high, since it is anticipated that a significant portion of the planning efforts devoted to the peripheral region of the EPZ may be eased. ,

Revised emergency planning criteria more closely linked to our best under- >

standing of accident risk can be expected to significantly enhance public  !

confidence, and lead to a more stable and efficient licensing procedure.

Tentative Agenda and Schedule: Theupdatedriskprofile(N' REG-1150)is  !

expected to be completed in FY 1986. The technical basis for change is i expected to be initiated in FY 1986 and rulemaking could be completed in

- late FY 1987.

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5. ACCIDENT CONSEQUENCES AND INDEMNIFICATION -

d; M. hackc round and Current Practice: Nuclear power plant licensees are required by

.- 10CFE50.54(W) to obtain the maximum onsite property damage insurance

'. s ' reasonably available, and by the Price-Anderson Act of 1957 as amended, are i? required to obtain the maximum liability insurance coverage available. Should an accident at any U. S. plant incur offsite liabilities in excess of that.

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. coverage, a retrospective premium would be paid by all Itcensees to create a fund to discharge this excess. At present, these two layers of offsite

, insurance provide $630 million of coverage, an amount that increases by $5 ,

million as each new nuclear power plant is licensed. The effective liability l? , insurance limit and the retrospective premiums are set by Congress, and do not

,., directly reflect an actuarial value of the indemnification provided. The Commission does, however, periodically provide recomendations and information

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concerning accident consequences, probabilities and liability limits to .

'J Congress. An example of such information is NURiG-0957. Noted therein is the

~ conclusion that the two layers of insurance should provide sufficient liability ,

protection for most accidents, but there remains a very low probability of high l consequence events that could result in public liability claims well in excess

of the available insurance. Whether source term research will change this risk ,

conclusion has not been establ.ished. , ,

In the event of an accident, the Comission must make a determination of I whether or not that accident was an "extraordinary nuclear occurrence" using criteria contained in 10CFR140. The radiological criteria in 10CFR140 ,

correspond to much lower severity levels than the dose guidelines of 10CFR100 and could, in theory, be etceeded by accidents not involving core damage. The accident at TMI-2 was, by 10CFR140 criteria, not an extraordinary nuclear ,

, occurrence. The Comission is considering revisions to the definition of an "extraordinary nuclear occurrence." , 3 Offsite risks of both human injuries and property damage are estimated by the  ;

, staff for ea

h plant for purposes of implementing the National Environmental ,

Policy Act (NEPA), and are reported in Draft and Final Environmental Statements prior to licensing. i Possibility for and Character of Changes: Amendments to the onsite property F insurance provisions of 10CFR50 are being considered. The current Price-i Anderson Act expires on August 1, 1987. In December,1983, the Comission  !

recomended that Congress extend the act and amend it to remove the liability

. limit. This last would be done by providing for retrospective premiums to be ,

i paid every year following an accident untti all claims are settled, and by '

', increasing the premium to $10 million per operating power reactor per year.

The majority of the Comission now believes that there should be a limit of i liability of at ledst $2.4 billion. It is possible that the future availability

. of the results of a thorough re-examination of the risks of severe accidents, using the new methodology, may affect the outcome of present or future

! Congressional deliberations of this matter; perhaps suggesting different l treatment of certain reactor / containment types.  ;

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n 5:1 The improved source terms will be used to compute more realistic offsite ,

y- consequences in future NEPA reviews.

O To determine what the impact of the new research would have on regulatory '

I requirements for onsite property insurance, two tasks would be undertaken. -

3 First, bases for the existing rule (NUREG/CR-2601) would be assessed in light

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. of the new information. Second, depending on the outcome of the first step, modifications to the existing rule would be proposed.

Vi a Benefits and Costs of Changes: Impacts on staff resources are expected to be

,i minimal. Since the risks of nuclear power are small, the expected costs and

., , benefits to individual utilities governed by the Price-Anderson Act are also small, actuartally. To be determined are licensee benefits from changes in '

, .? onsits or offsite insurance premiums. In the event of an accident, however, "J the difference between limited and unlimited indemnification could be immense 46 to the affected utility. Impacts on public confidence are unknown.

3

Tentative Agenda and Schedule: The staff will prepare a report to the Congress -

!. providing an assessment of the relationship of revised source terms on the

• Price-Anderson limit of liability. The report will be completed in late FY86 or early FY87, well before the expiration of the present act. .,

The new source term information will be used in preparing Draft Environmental

. Statements beginning in early FY86. -

6. AIR FILTRATION AND OTHER FISSION PRODUCT ATTENUATION ~ SYSTEMS Background and Curr'ent Practice: Engineered sa'fety features provided to mitigate accidental releases of fission products are reviewed and tested primarily for effectiveness against iodine. Their performance requirements are l generally derived from the TID-14844 assumptions. Guidance is provided through

"- several Regulatory Guides such as 1.3, 1.4 and 1.52, and the Standard Review s, Plan. Such systems include (1) containment spray systers, (2) recirculating

- air filters within containments, (3) control room habitability system air

,, filters,and(4)filteredbuildingexhausts. Radiological performance criteria

j. are based upon 10CFR100 dose guidelines for offsite exposures, and General f Design Criteria 19 (10CFR50, Appendix A) dose guidelines for control room .

l: operators. Since accidentally released iodine is presently assumed to be 91% i in the form of molecular vapor (I ). 5% in the form of particulate iodine, and 4%organiciodides,mitigativefe$turesgenerally(.ontainasprayadditive t g1 and/or charcoal impregnate intended to optimize the retention of iodine.  !

i) -

Many PWR containment spray systems employ caustic additives for direct addition  !

s to spray water, such as hydrazine or sodium hydroxide, to enhance removal- i

~; of elemental iodine. These systems are credited with such removal during l safety reviews BWR suppression pools and BWR containment sprays, which do not

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i, employ additive systems, presently are not credited with fission product L -

removal or retention even though they are capable of absorbing elemental iodine to some degree, t

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.y as d.i As an inducement to design rapidly acting systems, iodine is assumed to be  ;

$ released simultaneously with accident initiation and to leak into the  :

VS environment untreated during the time necessary for the filters to achieve full ,

P capacity. Often, large fractions of the astimated offsite thyroid doses are . .

';/3 due to the iodine releases postulated during the first few minutes of the '

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3 Possibility for and Character of Change: Present fission product mitigation C/ systems, as discussed above, are optimized for the removal r.d retention of ,

elemental and organic iodine. Source term research indicates that the todine

.. accidentally released in core melt accidents will appear no sooner than L. minutes after accident initiation and is likely to be present primarily in aerosol form along with aerosols of other volatile fission products. Less

,.T volatile fission products and other aerosols may be released later in core

,' melt accidents, k

"' The recireviation air filters within some containments contain particulate C filters specifically designed to remove aerosols. The ability of the 7 T. particulate filters to function in a revised source term environment should be analyzed.

Engineered Safety Feature filter systems outside containment 'are less'likely to be required to function in an environment loaded with aerosols, and. a .

change in source terms is less likely to impact their performance. ,

Since the effectiveness of present filter systems against all but high density i mixes of aerosols is believed to be high, and the same filter systems provide for fission product attenuation in accidents of lesser consequences in which -

iodine as a vapor (elemental) may be present, no short-term changes with regard to existing filter systems are anticipated. Changes in filter system  !

criteria indicated by the source term research are to be incorporated in  :

Regulatory Guide 1.52. High density aerosols and filter system bypass will also be considered. Backfitting of control rooms for removal of accidentally l released elemental iodine is to be delayed under a staff proposal until such

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time as guidance for conditions covering the range of accidents from those involving coolant releases to fuel melting are fully assessed. This staff proposal is presently being processed in the form of generic letter to .

Itcensees and applicants with regard to post-TMI backftiting requirements for control room operator protection. ,

Spray systems are known to be effective in removal of aerosols and additives contribute little to their already high aerosol removal capability. Such

- additives, however, are potentially important in post-accident pH control of L .. -

water that accumulates in plant sumps. At present there also exists a plant l l,- operational disincentive to have the spray system automatically actuated, lest  :

/ a corrosive additive ham equipment or personnel during maintenance  !

. activities, or in the event of an inadvertent spray actuation. The staff concludes, on the basis of the above factors, that staff guidance for injection of spray additives should be eliminated. To effect this change the l

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staff is preparing a draft Branch Technical Position and a generic letter to ,

% licensees and applicants. Credit should also be extended to BWR sprays and Mi suppression pools for fission product removal or retention in safety 1 GS analyses. These changes will a identified in modifications to the Standard .

..ig Review Plan. -

'! Cost and Benefits of Change: A lar e direct benefit to most PWR licensees is M- anticipated as a result of disconti uing the injection of very caustic spray ff ' additives. This is due primarily to reduced concern regarding the adverse effects of an inadvertent spray system actuation.

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A moderate, indirect benefit to BWR licensees is anticipated as a result of  ;

the granting of fission product removal credit for BWR suppression pools and

% sprays in treatments of design basis accidents and related perfomance

" requirements for some-Engineered Safety Features. Moderate staff costs are i '

expected to accumulate from developing and implementing changes to Regulatory ,

Guides and the Standard Review Plan.

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With respect to safety related filtration systems, it is expected that the >

source term may only result in changes for filtration systems inside containment with respect to particulates. Yet to be reassessed are the

-_ criteria for filtration systems outside containment, including those for  ;

operator protection in control room habitability. systems. .

Tentative Agenda and Schedule: These changes are expected to be initiated in ,

FY86 and completed in FY87.

7. ACCIDENT MONITORING AND MANAGEMENT. AND ONSITE AND OFFSITE INSTRUMENTATION .

. Background and Current Practice: All plants are required to provide

. Instrumentation to monitor plant variables and systems during and following an J. accident (TMIActionPlanII.F.1.2). This instrumentation is calibrated and l -

qualified for conditions estimated from the fission product assumptions of 4 F TID-14844 and the peak accident temperature and pressure profiles for other design basis accident assumptions. Under Regulatory Guide 1.97, some instru-p ments considered necessary to follow reactor accidents are qualified and i calibrated to a still harsher radiation environment. Additionally, research on s

' , ' accident management is under way that may improve guidance for prevention and

. mitigation, i Instrumentation is also required for offsite monitoring of both routine and l

. accidental releases. Offsite instrumentation includes fixed themo- ,

[C luminescent detectors (TLD) at an array of locations surrounding the site, <

3 portable radiation measuring equipment, and portable air samplers. .

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. Possibility for and Character of Change: The instrumentation required to f

., assess plant conditions during and following an accident is qualified to an  :

environment consisting of the non-mechanistic radiation doses indicated by i TID-14844 release assumptions (except for a small number, as indicated above,  ;

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N/ which are qualified to a more stringent environment) combined with temperature .

and pressure profiles calculated for pipe break accidents not involving a core melt. Source tem research may require changes in the range of parameters monitored by these instruments.

W The source term related variables that could enable an operator to follow the h course of accidents and potentially prevent or mitigate releases should be it re-evaluated. Given this identification, it would be possible to evaluate the risk significance and benefits for instrumentation needed to measure the (S: identified variables and their capability of surviving the environment to which

- they might be subjected for the period of time for which they must operate.

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Furthermore, emerging related research on accident management may provide guidance on the types of instrumentation most useful in preventing accidents, g for following the course of accidents, and for mitigating. releases.

. The radiological criteria in 10CFR140 by which an extraordinary nuclear occurrence is to be determined, and specific offsite indemnification criteria are to be invoked, are given in terms of human organ doses and surface e contamination densities by specific classes of radioisotopes. These measure, ments, however, are not simply and unambiguously made by the TLD's and portable

, equipment presently available. This difficulty is exacerbated by the new source term information for many kinds of accidents, which decrease the likely offsite dose contributions expected from noble gases (by virtue of potential delayedcontainmentfailure)andiodine. Offsite monitoring requirements should be reevaluated using new source term information to assure an adequate diagnostic capability. It may prove riecessary either to recast the radio-logical criteria of 10CFR140 in terms more directly related to measurable quantities, or to require specific radio-assay equipment in offsite monitoring programs.

Costs and Benefits of Change: The expense of requalifying instruments,

.: changing their monitoring range, and their supporting electrical equipment could be large, and would only be justified in those instances in which the assured operation of particular instruments could be shown to significantly reduce the likelihood of an accident and its consequences.

Offsite portable equipment is not a major expense, but a large intangible benefit exists in adequately quantifying radioactive releases, both routine and accidental, in order to adequately evaluate estimated health effects.

Tentative Agenda and Schedule: The staff will identify variables that are of c- risk significance in preventing accidents from progressing in severity or in A mitigating those that are not adequately covered by the guidance in Regulatory Guide 1.97. The identification of changes in criteria is expected to be completed by the end of FYB6.

[, Offsite monitoring will also be re-assessed by the close of FY86, and any

., necessary revisions to Regulatory Guide 1.101 proposed for completion in FY87.

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8. OFFSITE CONTAMINATION AND RECOVERY '

1. Background and Current Practice: Considerations of offsite contamination and i recovery resulting from an accident is applied in a number of areas of i's regulatory oversight. As part of its assessment of standard plant risks and d for NEPA purposes, applicants, licensees and the staff analyze the offsite J.d effects of severe accidents, including health effects, costs of contaminated i real property, crop interdiction, evacuation and decontamination costs. In

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. one case, for the GESSAR II review of severe accidents, the staff has used the

. new source term information. In other cases such assessments have used source

' terms derived from WASH-1400. Further, WASH-1400 source terms have also been a

)! basis of risk judgments on activities such as possible use of severe accident l

• mitigationfeatures(e.g., filtered-ventedcontainments)toreducetheimpact I

i, of offsite contamination. Lastly, while not directly applied in the regulatory

">.- arena, the . judgments made using WASH-1400 source tenns regarding offsite

_. contamination levels may have had a significant influence upon public attitudes

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toward nuclear power.

e- Probability for and Character of Change: Revised source term research is

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likely to change estimates of both the absolute as well as the relative amounts of fission products released.' Although research is incomplete, the results appear to be generally lower in absolute quantities for some accident -

sequences, and highly dependent upon plant design. Changes in several areas

! may be anticipated. ,

In forthcoming standard plant and NEPA evaluations the staff expects to examine'

accident risks as well as offsite contamination and recovery costs utilizing the i new information. It is anticipated that the new source term results will

( generally show a lowar intensity of offsite contamination than would be .

l predicted using WASH-1400 assumptions, but with potentially larger areas.

These evaluations are directly related to the cost-effectiveness of a number i i of engineered safety features that have been proposed to mitigate the  !

consequences of a severe accident, such as filtered-vented containments.  !

Costs and Benefits of Change: A moderate staff effort is anticipated in i utilizing revised source term infonnation for forthcoming standard plant reviews. This will be principally in the area of obtain'ng revised source tenns to provide reasonable risk perspectives. A moderate staff effort is i expected in the area of assessment of engineered safety features in light of r l, revised source tem information. Some research and scoping studies have been  !

i. perfonned in the past. These have indicated that the addition of many types of  ;

,e engineered safety features is not cost effective from'a risk perspective. Work u

on standardized plants is expected to demonstrate this conclusion.

( Licensees are not expected to benefit directly from the anticipated changes in this' area. The benefits are expected to flow from research and scoping studies on proposed engineered safety features for severe accident mitigation which are  !

' not expected to result in any direct costs to licensees and applicants, but  !

J' will likely build confidence in existing designs. l

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L.; Overall, the benefits in this area are expected to be primarily intangible j ones, largely in the area of increased public confidence.

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• Tentative Agenda and Schedule: Assessment of proposed ESF's is being pursued as part of the severe Accident Research Program and standard plant reviews. -

e Together with staff interaction with the industry group (IDCOR), evaluations

,% expected in FYB6 and FY87 are expected to demonstrate generally reduced risks pr and lowered requirements for additional ESF's for standard plants.

9. SAFETY ISSUE EVALUATIONS 9 Background and Current practice: This area includes both the evaluations of generic safety issues from their identification to their resolution (including f backfitevaluations),andsafetyevaluationsforlicensingandlicensing amendments.

) Each newly raised generic safety issue is assessed by the staff in a two-step

, process. In the first step, the staff establishes the potential importance of the new safety issue relative to all others, in order to assign it a priority in the competition for limited resources. This prioritization is a systematic application of PRA methods to estimate two indices of safety importance:

o Risk Importance - an assessment of the increase in societal risk posed by

the generic issue as indicated by the estimated dose to the surrounding population out to 50 miles from the plant; and t o Value/ Impactthe issue, i.e., - aratio measure of risk of the the cost effectiveness reduction of resolving to the total cost to (the safety i industry and the NRC) involved in developing and implementing the mode of resolution, i The indices are currently evaluated using source tenns derived from WASH-1400.

E Generic safety issues assigned a high or medium priority in the use of staff L resources enter the second step of the process. A more detailed evaluation is then performed, in which risks under existing conditions are compared to risks under the improved conditions that would be brought about by increasing l, regulatory requirements to resolve the generic safety issue. Among the factors p considered in this second step are the potential reduction in risk to the public and the potential impact on the radiological exposure to plant personnel associated with the proposed new requirements or backfit. Currently, the probabilistic estimate of public risk uses accident source terms derived from '

R ,." WASH-1400. On the other hand, occupational exposures are estimated in a R- relatively direct manner based on measured onsite radiation levels and expected '

p- times of exposure.

U Safety evaluations in support of licensing and technical specification changes

'1

f. - are based in large measure on stylized accident calculations derived from -

m TID-14844 and related assumptions. These calculations are used in safety -

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3 evaluations for construction permits, operating license's and license amend- i t.9 ments. The calculational methodolor is described in TID-14844, various

'p.4 Regulatory Guides and the Standard Review Plan. The methodology includes not M only assumptions of fission product releases for a number of design bases 6: accidents, but other assumptions relating to the transport and deposition of s releases and their biological uptake some of which have not been updated since 1962. All such assumptions were deterministically designed to be conservative

.: in combination, but not individually to envelop all conditions. .

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Possibility for and Character of Change: The new source term methodology

'7 . can be used to estimate accident consequences and risks. Such estimates used  ;

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in prioritizing generic issues will utilize the new research. In many cases >

', . involving backfitting considerations, the assessed value/ impact could differ ,

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enough to alter conclusions. These may be different for different plant types  ;

or groups. The staff anticipates revising the indices by which the safety fj importance of generic issues are evaluated to agree with the insights gained

from the recent source term research, but because the overall methodology 5 allows for a rough screening of the relative importance of the issues, no ,

J ,' reprioritization is planned.

In view of the source term research, and related information developed since TID-14844 was published, the staff will reassess its methodology for assessin the radiological consequences of design basis accidents. Four alternat ves will be considered as follows:

a. Evaluate whether the existing methodology is sufficiently conservative to maintain the status quo, including the use of TID-14844 assumptions, for

, the suite of design bases accidents presently assessed.

I b. Determine whether modest changes to present assumptions, such as the i

fission product release assumptions in TID-14844, can be restated for all >

plants, or groups of plants, to provida a suitable level of safety for ,

. future regulatory analyses. Such restatements can.be made in the form of '

Regulatory Guide and Standard Review Plan changes.

• j

!-. i

c. Restate all design bases accident assumptions in one of two ways; <

h (1) recast design basis accident assumptions deterministically using the . t same kinds of dose assessments as presently employed;.or

. (2) do away with design basis accident dose calculations in favor of 1

performance criteria for engineered safety features. '

u. . i Q

Rulemaking and Regulatory Guide and Standard Review Plan changes may be  ;

',' necessary under this alternative. '

(N.' d. Establish a risk based evaluation methodology using maturing probabilistic i * ~l .

risk assessment techniques. Rulemaking and Regulatory Guide and Standard y

Review Plan changes may also be necessary under this alternative.  ;

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16 ENCLOSURE 1 P.' Costs and Benefits of Chenge: Moderate staff effort is expected to revise the. .

9 evaluation methodology of generic safety issues using new source term b,2 3 information. Benefits are expected to be high since the resolution of an issue will be in keeping with the latest research data and, therefore, will emphasize more accurately areas of risk significance. This change may result in a N'G-g general. risk reduction for many generic issues.

ff

%c A modest effort is also expected to identify and implement changes to the present staff design basis accident evaluation methodology. Benefits could be g modest. .

Tentative Agenda and Schedule: Revision of the methodology for evaluating generic safety issues will begin in FY85 and be completed in FY86.

Q N: Consideration of alternatives to the present design basis accident methodology 3 is expected to be completed in FY86 and the results implemented in FY87.

g.  :

C 10. SITING r, '

4 Background and Current Practice: Siting criteria in 10CFR100 include several tests in which it must be demonstrated that (a) the site possesses certain ,

characteristics; (b) the plant-site combination meets certain criteria; and y (c) the site is located sufficiently far from population centers (remote r siting). Each site must have an exclusion area within which an applicant has authority to, determine all activities. Beyond the exclusion ar.ea lies the low

. population zone (LPZ) within which the total number and density of residents ,

,. must be such that there is a reasonable probability that protective measures p could be taken in their behalf in the event of an accident. Finally, Part 100 t

requires that the distance from the reactor to the nearest population center (of about 25,000 or more residents) must be at least one and one-third times a '. the distance to the outer boundary of the LPZ. The distances to the exclusion

[ area boundary, the LPZ outer boundary and the population center are not numerically fixed, but depend upon the plant characteristi,:s, including its o maximum full power fission product inventory, and the complement and

• performance of certain engineered safety features.

To test whether the plant-site combination meets the requirements of Part 100,

, a hypothetical core melt accident (TID-14844) is postulated involving the r instantaneous release of 100% of the core inventory o noble. gases and 50% of '

M the radiciodines into the containment. Half the iodines released are assumed to plate-out upon interior surfaces, while the remaining 25% is available for s leakage (primarily as elemental iodine). The containment is assumed to remain

y intact, but is presumed to leak at its design basis leak rate. The perfonnance

v. of fission product mitigating features (e.g., sprays, filters) are assessed in .

W. a stylized fashion to estimate thyroid and whole body doses unlikely to be '

u- l exceeded in hypothetical individuals located at the exclusion area boundary and N the LPZ outer boundary for specified time periods. The plant-site combination

u. . . is determined to be acceptable if the calculated doses do not exceed the i guideline values given in Part 100. It is clear that current licensing

^

!; . practice involves a close coupling of the site and the plant design. Also, the  ;

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d 17 ENCLO5URE 1 5::

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9 thyroid dose calculation, driven by assumptions of the release of elemental ,

p. .

iodine, is usually the limiting dose in detencining the acceptability of the exclusion area boundary, the low population zone, and performance requi'rements

$>; of certain fission, product mitigating features. ,

W

' No dose calculations are performed for individuals located at the population center. However, in the statement of considerations accompanying the issue of W Part 100, it was noted that the population center distance requirement was C.; added to provide for protection against excessive accident exposure doses to t people in large centers, since accidents greater than the hypothetical accident is postulated for siting purposes was considered conceivable, although highly *

.A . improbable. This statement was recognition of the possibility of accidents y involving greater consequences than design basis accidents and of their

.y importance in titing considerations. Since publication of the Reactor Safety fe Study (WASH-1400) in 1975, there has been a general recognition that although M probably small, considerations, public accident risks are dominated by such accident y

w .

/ These siting criteria, promulgated in 1962, led to a significant improvement in

.. fission product mitigating engineered safety features in the late 1960's and

early 1970's. In response to observations that 10CFR100 did not preclude sites with very small exclusion area boundaries or in relatively densely v populated regions, the staff proposed in 1975 (Regulatory Guide 4.7) that l . exclusion area distances should be about 0.4 miles or greater and that the

, average population density for the circular region surrounding a site should have no more than about 500 people per square mile within a distance of 30 miles. The effect of this guide is to suggest a minimum stand-off distance from reactor sites to large population centers (for example, a reactor should be located about 25 miles away from a city of 1,000,000 people). Sites where d

these values are exceeded are not forbidden, but should be shown to possess

. superior features in other respects to offset the disadvantage of high popula- .

tion. About 90 percent of the 75 U.S. power reactor sites meet the density

!? criteria of the Guide; those that do not were reviewed and approved prior to 1975. The criteria of Regulatory Guide 4.7 were intended to provide a

'c reasonable degree of separation from large population centers while maintaining 0 a good availability of land area for potential future sites. However, no y explicit consideration of severe accidents was employed.

l. In1979,astaffevaluationofsitingpolicyandpractice(NUREG-0625) reconsnended, among other things, an explicit consideration of severe accidents i'e[

in siting, and a decoupling of plant design and siting requirements. The T Comission initiated rulemaking in this area in 1980, but suspended it about a

( ..' year later pending a reevaluation of accident source tenns as well as an y evaluation of the proposed safety goal.

I s.

. Possibility for and Character of Changes: Changes in siting could come about

[A primarily in two areas. Thesearein(1)evaluationofcertaindesignbasis g accidents usedtin licensing evaluations-and technical specifications, and -

g (2) consideration of severe accidents. .

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.18 ENCLOSURE 1 .

V a. The present design basis accident postulated for testing combinations of I

% ESF and site characteristics am based on a source term that may not -

h d -

provide an adequate characterization of accident consequences. Among these are an undue attention to iodine, particularly in the elemental form, and -

1. 5 neglect of other fission products of importance such as cesium. The possibilities for change with regard to design basis accidents are; W@

R (1) Develop revised design basis accident assumptions for siting and

.i evaluation procedures for engineered safety feature perfonnance that-are in accord with the evolving understanding of fission product f behavior under degraded core conditions. '

fb *

(2) Eliminate design basis accident radiological evaluations related to i site / plant adequacy entirely by specifying a minimum set of required p? engineered safety features together with their perfonnance criteria,

1. plusaminimumsetofsitecharacteristics(e.g.,distancesto i exclusion area boundary, LPZ and population center). This -

1. possibility is the essence of the recommendations in NUREG-0625. .

b. Siting criteria presently contain no explicit considerations of severe l

<(~ accidents, as noted above. Comission policy,10CFR100, and present staff practice (via Regulatory Guide 4.7) does encourage siting away from densely populated centers, but the present population density l J. criteria have no clear link to severe accident risk. It should be

. noted, however, that more than 20 site-specific Environmental Impact 3

Statements (EIS) have been performed by the staff which have included c an explicit discussion of severe accident risks. These have shown i a' the risks to be low for all the sites analyzed. In addition, the '

! Comission special proceeding for the Indian Point site, the highest l N population density reactor site, also explicitly considered severe j

.. accident risks and concluded these to be low. Based upon evaluations .

4J, of low accident risks for present sites using WASH-1400 source tenns,  !

$ together with the fact that there appears to be little incentive for  :

siting in more populous areas, the staff anticipates that a revision P' of siting criteria employing new source tenns would represent no i major changes from present staff practice with respect to site '

j selection by applicants.  !

Costs and Benefits of Change: Revision of siting criteria is expected to l

,, require a low to moderate effort in regard to staff resources. Most of this  ;

P. - would be in evaluating the risks associated with the proposed criteria and ,

f alternative approaches, including an evaluation with regard to the Safety Goal l 2 criteria..

i.Y Benefits in terms of direct regulatory relief are expected to be low, since, as  !

!W mentioned earlier, the staff anticipates no major departures from present staff  !

t practice and there is little incentive for more populous siting. Intangible .

7 benefits' in terms of contributing to the Comission's policy on preapproval of  ;

y- plant sites and enhancement of public confidence are expected to be high,-

l

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'i!.i 19 ENCLOSURE 1 n,! ^

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'

?jY however. The benefits also include the potential for a more stable and

~'

,, j efficient licensing process.

r.. ;

i,) Tentative Agenda and Schedule: Since rulemaking in this area had been r,q: initiated In 1980 and suspended'about a year later, some of the technical work, CJ especially in the area of land availability, is considered to be still applicable. It is estimated that rulemaking could be reactivated in FY86

){/!;.

and completed in early FY87. ,

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g ENCLOSURE 2 t

.I h PRELIMINARY BENEFIT-COST SUW4ARY OF AREAS N TARGETED FOR SOURCE TERM RELATED CHANGES .

W p CHANGE IN A REGULATORY REGULATORY IMPLEMENTATION AREA COSTS BENEFITS REQUIREMENTS

• TARGET

[g u .

p 1 IDCOR-NRC O Staff Search TO BE DETERMINED

, For Risk Outliers Es 2 Containment

< Perfonnance

, Near Tem Low High D/I 1-2 years, Future Unknown Unknown D/I U 3 Equipment Moderate Moderate Qualification to High to High D/I 2-3 years 4 Emergency Planning Moderate High D 1-2 years 5 Accident Consequences Low Low U 1-2 years

& Indemnification 6 Air Filtration Low High U 1-2 years

& Other Fission L- Product Attenu-ation Methods 7 Accident High High U 1-2 years Monitoring & ,

,- Management Onsite

& Offsite Instru-f ,. mentation lt

- 8 Offsite Con-

-/ tamination & Low High D 1-3 years

,. Recovery 9 Safety Issue Moderate High U 1-2 years j Evaluations 10 Siting Low to Moderate Moderate U 1-2 years

• Decrease (D), Unknown (U), Increase (I)

• A

h,,

t L. SorrW8 l

CUARTERLY SOURCE TERM BRIEFING FOR THE COMMISSION NOVEMBER 4, 1985 9

i e

e A,

27 17 J

m v.: i *. ,

TABLE OF CONTENTS A. SEVERE ACCIDENT IMPLEMENTATION PROGRAM AND SOURCE TERM RELATED REGULATORY CHANGES (NRR) (PAGES 1 - 14)

B. STATUS OF NUREG-1150 (PAGES B-1 TO B-11)

C, PUBLIC COMMENTS ON NUREG-0956 b

2:

D. STATUS OF SOURCE TERM CODE PACKAGE i .

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SEVERE ACCIDENT IMPLEMENTATION PROGRAM AND SOURCE TERM RELATED REGULATORY CHANGES COMMISSION BRIEFING 4

NOVEMBER 4, 1985 b

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T. P. SPEls/NRR X27517 O

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-SEVERE ACCIDENT POLICY IMPLEMENTATION PROGRAM J

OVERVIEW  ;

i PROCESS OF IMPLEMENTATION c I

i i

EXPECTED OUTC.0MES POTENTIAL' ISSUES SOURCE TERM RELATED CHANGES .

4 PROPOSED SCHEDULE b

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E

SEVERE ACCIDENT POLICY IMPLEMENTATION PLAN I,

SUMMARY

OF THE IMPLEMENTATION PROGRAM SEVERE ACCIDENT POLICY STATEMENT - ACTION ITEMS GUIDELINES AND PROCEDURAL CRITERIA FOR THE EXAMINATION OF INDIVIDUAL PLANTS INDUSTRY PARTICIPATION EXPECTED ACCOMPLISHMENTS POTENTIAL ISSUES ,

II, SOURCE TERM RELATED CHANGES IN RULES AND REGULATORY PRACTICES DEVELOPMENT OF NEW SOURCE TERMS AND INITIATION OF RULE CHANGES t '

INFORMATION NEEDED TO INITIATE SOURCE TERM RELATED

~

CHANGES POTENTIAL CHANGES IN RULES AND REGULATORY PRACTICES III. PROPOSED SCHEDULE

1. 4 i

---

_j

.,g SEVERE ACCIDENT POLICY STATEMENT - ACTION ITEMS POLICY STATEMENT NEW APPLICATIONS EXISTING PLANTS GUIDANCE ON THE ROLE SYSTEMATIC APPROACH OF PRAs FOR THE EXAMINATION OF INDIVIDUAL PLANTS

~

^

PERFORMANCE CRITERIA IMPLEMENT MODIFICATION FOR CONTAINMENT THROUGH BACKFIT POLICY SYSTEMS CHANGES IN RULES AND REGULATORY PRACTICES, AS MEEDED 3

, ~.

GUIDELINES AND PROCEDURAL CRITERIA FOR THE EXAMINATION OF INDIVIDUAL PLANTS SEVERE ACCIDENT SAFETY ANALYSES ARE BEING PERFORMED FOR SIX REFERENCE PLANTS, RESULTS TO BE PUBLISHED IN AUGUST, 1986 FOR COMMENT AND PEER REVIEW (NUREG-1150),

UNCERTAINTIES WILL BE ASSESSED AND QUANT.IFIED.

THE PURPOSE OF THE REFERENCE PLANT EVALUATIONS IS TWOFOLD:

- EVALUATE PERFORMANCE WITH RESPECT TO SEVERE ACCIDENTS

,- DEVELOP GUIDANCE AND CRITERIA FOR THE SYSTEMATIC EXAMINATION OF INDIVIDUAL PLANTS FOR PLANT SPECIFIC ACCIDENT VULNERABILITIES (SUBSTANTIAL INPUT FROM IDCOR PROGRAM),

THE GUIDELINES WILL SPECIFY:

- CONDITIONS INDIVIDUAL PLANTS MUST MEET TO "AD0PT" THE REFERENCE PLANT ANALYSIS,

-o

? - THE EXTENT OF A PLANT DEPENDCNT EXAMINATION LICENSEES NEED TO CONDUCT, AND

~

- ACCEPTABLE METHODOLOGY TO SEARCH FOR PLANT SPECIFIC

- VULNERABILITIES, e

~

INDUSTRY PARTICIPATION (IDCOR)

IDCOR HAS ANALYZED FOUR OF THE SIX REFERENCE PLANTS AND PRESENTED THE ANALYSES TO NRC. THE IDCOR ANALYSES ARE-AN INTEGRAL PART OF THE REFERENCE PLANT EVALUATIONS AND THE DEVELOPMENT OF THE GUIDELINES, BASED ON THE IDCOR PRESENTATIONS AND UNDERSTANDING GAINED FROM THE SEVERE ACCIDENT RESEARCH PROGRAM, NINETEEN TECHNICAL ISSUES WERE IDENTIFIED WHICH WERE TREATED DIFFERENTLY BY THE TWO PARTIES AND WERE JUDGED TO HAVE A SIGNIFICANT EFFECT ON THE RESULTS. THE RESOLUTICN OF

~ THESE ISSUES IS IN PROGRESS, THE IDCOR REFERENCE PLANT ANALYSES HAVE TWO SHORTCOMINGS:

- THE UNCERTAINTIES ASSOCIATED WITH THE ANALYSES HAVE NOT YET BEEN EVALUATED.

l - SEISMIC EVENTS AND OTHER EXTERNAL EVENTS HAVE NOT BEEN

! CONSIDERED, DECISION IS PENDING, IDCOR TS CURRENTLY DEVELOPING METHODOLOGY FOR THE EXAMINATION OF INDIVIDUAL PLANTS, AS PART OF THE l DEVELOPMENT EFFORT, THE METHODOLOGY WILL BE APPLIED TO THE FOUR IDCOR REFERENCE PLANTS AND THREE ADDITIONAL PLANTS CE, A B&W DESIGN, AND A BWR-MARK II). THE METHODOLOGY WILL BE SUBMITTED TO NRC FOR REVIEW, NRC APPROVAL WILL BE ISSUED AS PART OF THE GUIDELINES, e

b 5

EXPECTED ACCOMPLISHMENTS PLANT SPECIFIC VULNERABILITIES WILL BE IDENTIFIED AND FIXED (BACKFIT RULE)

• IF GENERIC VUL'!ERABILITIES ARE IDENTIFIED, APPROPRI ATE

' DESIGN AND/0R OPERATIONAL CHANGES WILL BE REQUIRED (RULEMAKING) ,

LESSONS LEARNED WILL HELP DEVELOPMENT OF IMPROVED DESIGNS WITH SAFETY BENEFITS b

5'

• A NEW,'MORE REALISTIC REGULATORY APPROACH ON SOURCE TERMS j

~

WILL BE PURSUED (SOURCE TERM RELATED CHANGES) i l

4 8

6 .

POTENTIAL ISSUES l l

LARGE DIFFERENCES COULD EXIST BETWEEN IDCOR AND NRC CALCULATIONS QUANTIFICATION OF THE UNCERTAINTIES COULD RUN INTO DIFFICULTIES '

COMPLETENESSOFPL%NTANALYSESAREIbQUESTIONBECAUSEO EXTERNAL EVENTS DEVELOPMENT OF PRACTICAL SOURCE TERMS IS A NON-TRIVIAL y ISSUE ,,

O 9

e 4

- 7 -

OUTSTANDING TECHNICAL ISSUES CORE MELT ISSUES

• CORE MELT PROGRESSION; IN-VESSEL HYDROGEN GENERATION, FISSION PRODUCT AND AEROSOL RELEASE RETENTION AND REVAPORIZATION OF FISSION PRODUCTS IN THE REACTOR-COOLANT SYSTEM CONTAINMENT ISSUES .

EX-VESSEL FISSION PRODUCT AND AEROSOL RELEASE AND DEPOSITION SCRUBBING EFFICIENCY OF SUPPRESSION POOLS AND ICE CONDENSERS 9

CONTAINMENT LOADJ (P/T)

CONTAINMENT FAILURE MODES EQUIPMENT PERFORMANCE IN SEVERE ACCIDENT ENVIRONMENT e

8

SOURCE TERM RELATED CHANGES IN RULES AND REGULATORY PRACTICES DEVELOPMENT OF NEW SOURCE J_@RM AND INITIATION OF RULE CHANGES SOURCE TERM RELATED REGULATORY REQUIREMENTS AND PRACTICES WERE REVIEWED. TWELVE AREAS WERE IDENTIFIED WHERE NEW SOURCE TERM KNOWLEDGE MAY LEAD TO CHANGES,

, .THE VARIOUS REGULATORY USES OF SOURCE TERMS COULD REQUIRE DIFFERENT FORMS OF SOURCE TERMS, NRC WILL EVALUATE AND GROUP THE TWELVE AREAS OF APPLICATIONS OF SOURCE TERMS, AND WILL DEVELOP PRACTICAL SOURCE TERMS FOR EACH GROUP, THE NEEDED FORMS OF SOURCE TERMS WILL BE DISCUSSED WITH THE NUCLEAR INDUSTRY (AIF),

- WE FORESEE, AT MOST, THREE FORMS OF SOURCE TERMS:

o T (1) DEIAILED SOURCE TERM CALCULATIONS FOR INDIVIDUAL PLANTS, (2) USE OF TABLES OR PROCEDURES APPLICABLE TO PLANT TYPES, AND (3) A SIMPLE, BOUNDING SOURCE TERM APPLICABLE TO ALL PLANTS, ,

l NRC IS PRESENTLY APPLYING THE DETAILED SOURCE TERM METHODOLOGY FOR THE ANALYSES OF THE SIX REFERENCE PLANTS (NUREG-1150), THE REFERENCE PLANT ANALYSES WILL PROVIDE SOME BASIS FOR ARRIVING AT SIMPLIFIED FORMS OF SOURCE I TERMS, CHANGES TO CURRENT REGULATIONS AND REGULATORY l

PRACTICES WILL BE INITIATED AS S00N AS AVAILABLE

.INFORMATION WARRANTS IT, s

. 9

l Information Needed to Initiate Source Term Related Changes ,

Understanding of Development of .

8'I***fc"p Pi a for Na Phenomena, and New Source Term Source Tarm P " ' e Detailed Calculations I l h*e*h do ogy for Example: Core Melt, but Source Term

• I*s or Pro ed" ** No Early Containment l , py 9 ,ce er J

Calculations Failure t r

.l I

f 1 f Changes in Rules and Regulatory Practices ,

a I

10

POTENTIAL CHANGES-IN RULES AND REGULATORY PRACTICES SHORT-TERM (T0 BE STARTED PRIOR TO ISSUANCE OF NUREG-1150)

REMOVE SPRAY ADDITIVES (PWRs)

CREDIT FOR FISSION PRODUCT SCRUBBING IN SUPPRESSION POOLS (BWRs)

REVISED TREATMENT OF SEVERE ACCIDENfS IN NEAR-TE ENVIRONMENTAL IMPACT STATEMENTS 6

Y ..

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POTENTIAL CHANGES IN RULES AND REGULATORY PRACTICES I

INTERMEDIATE TERM (BACKGROUND EFFORT PROCEEDS IN PARALLEL Wi1H NUREG-1150 EFFORTS RULEMAKING BEGINS AFTER ISSUANCE OF NUREG-1150)

EMERGENCY PLANNING - ONSITE PLANNING, OFFSITE EMERGENCY PLANNING ZONES (EPZ), GRADED RESPONSE

• - CONTAINMENT LEAK RATES - LEAK RATE TESTING, UNDETECTED BREACH OF CONTAINMENT INTEGRITY CONTROL ROOM HABITABILITY - FILTRATION AND LEAK-TIGHTNESS REQUIREMENTS S

ENVIRONMENTAL QUALIFICATION OF EQUIPMENT ACCIDENT INDEMNIFICATION - POTENTIAL RENEWAL OF PRICE-ANDERSON SAFETY ISSUE EVALUATION - PRIORITIZATION OF ISSUES USING NEW SOURCE TERMS e

e 12 -

POTENTIAL CHANGES IN RULES AND REGULATORY PRACTICES LONG-TERM OFFSITE CONTAMINATION AND REC 0VERY SITING - EXPLICIT CONSIDERATION OF SEVERE ACCIDENTS IN SITING ACCIDENT MONITORING AND MANAGEMENT - ONSITE AND OFFSITE

~ INSTRUMENTATION l

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Proposed Implementation Schedule

~

t .

Tasks 1985 1986 ' 1987 1988 l Method Development for Analysis -- = = = = = = = = == = = = = = = '---=

l Reference Plant Analyses (NUREG-ll50) --- -

l l

i Method Development for Individual Plant -

e-o ===

Examination (1DCOR) ,

Development of Guidelines and Criteria for en- ,---

Individual Plant Examination i '

Examination of Individual Plants (Licensees ) .

J Changes in Rules and Regulatory Practices Legend:

/ .

Execution of Task

--- Review, Public Camments..

) ,

improvements I

14

STATUS OF NUREG-1150 9

COMMISSION BRIEFING S

r NOVEMBER 5, 1985 e

6

NUREG-1150 .

o RISK PERSPECTIVES AND PROFILES ON SIX REFERENCE PLANTS SURRY PEACH BOTTOM SEQUOYAH GRAND GULF ZION LASALLE o RISK REDUCTION POTENTIAL

- GENERIC NUREG-0900 FIXES (E.G., FILTERED VENTED CONTAINMENT, ADD-0N DECAY HEAT h- REMOVAL, ADDITIONAL CONTAINMENT HEAT REMOVAL, ETC.)

l PLANT-SPECIFIC MODIFICATIONS SUGGESTED BY ACCIDENT SEQUENCE REBASELINING AND CONTAINMENT TREE INSIGHTS o DISPLAY AND CONSIDERATION OF IMPORTANT UNCERTAINTIES o INSIGHTS ON USE OF RISK INFORMATION FOR PLANT-SPECIFIC AND GENERIC REGULATORY APPLICATIONS 32

4 NEED FOR EXPANDED PROGRAM o POTENTIAL THAT DIFFERENT ACCIDENT SEQUENCES MAY BE RISK-IMPORTANT INCREASED THE IMPORTANCE OF UPDATED FREQUENCY INFORMATION, REBASELINING ALLOWS CONSIDERATION OF:

- IMPROVED INSIGHTS FROM PRAs IN RECENT YEARS (E.G.,

COMMON-CAUSE FAILURES, PLANT-SPECIFIC DATA, HUMAN RELIABILITY) 5 PLANT AND PROCEDURE IMPROVEMENTS SINCE TMI b

?

o NEED FOR MORE SOURCE TERM CODE RUNS--BMI-2104 RESULTS INDICATED THAT SOURCE TERMS FOR SOME ACCIDENT SEQUENCES l COULD BE LARGE, AND IT IS DIFFICULT TO EXTRAPOLATE EXISTING RUNS REALISTICALLY TO RISK-lMPORTANT SEQUENCES e

4

,av- y w.-. .... ., y. ...., ,

10/31/85 .-

. e er (gj3g)) ~

FINAL SARRP SCHEDULE FOR_NUREG-1150 AND NRR SUPPORT SURRY PEACH BOTTOM SEQUOYAH GRAND GULF ZION

• LASAbl ACTIVITY

1. ACCIDENT SEQUENCE C C C C 12/19 INITIAL INPUT C ,

2. SOURCE TERM C C C C 1/23 1

BINNING C

) 3. NUMBER OF SOURCE

~8 3 N6- N6 TERM CODE RUNS N3 6 C C 11/15 11/15 12/31 3/20/8(

4. SOURCE TERM CODE RUNS (EXCEPT 1)

C C 11/22 11/25 1/15/86 4/14/81

5. RELEASE CHARACTER-ISTICS (EXCEPT 1)

C 11/15 12/10 12/11 1/31/86 4/28/8'

6. CONSEQUENCE CALC. '

(EXCEPT 1) 1 7. REFINE ACCIDENT 11/15 12/13 N/A 2/27/8-i SEQUENCES C 11/11

8. CONTAINMENT TREES 12/3 1/20/86 1/31/86 5/28/8 DRAFT REPORT 11/15 12/9

9. BASELINE RISK 12/18 1/7/86 2/15/86 5/26/8 11/19 11/25 CALCULATION

10. RISK / RISK RED. 3/14/86** 3/28/86** 6/9/86 TABLES 12/9 1/20/86 1/21/86

11. RISK / RISK RED. DRAFT 4/15/86 4/30/86 7/7/86 DETAILED RPT. 2/26/86 " 4/6/86 " 4/1/86 "

(d12. RISK /RISKRED. FINAL 6/1'/86 5/27/86 6/10/86 6/30/86 9/1/86 y DETAILED RPT. 4/23/86

• TENTATIVE SCHEDULE "INCLilDES FINAL SENSITIVITY ANALYSES AND UNCERTAINTY RANGES

~

PRESENT SCHEDULE (COMPARED TO 6/27/85)

PLANT RISK / RISK REDUCTION TABLES AND EXPLANATORY TEXT 6/27/85 10/31/85 SURRY 10/15/85 12/9/85 PEACH BOTTOM 11/30/85 1/20/86 SEQUOYAH 12/30/85 1/21/86 GRAND GULF 1/30/86 3/14/86*

r 3/28/86*

~

ZION" 2/28/86 LASALLE N/A 6/10/86*

i NUREG-1150 MID SUMMER OF 1986 (N0 CHANGE)

• lNCLUDES FINAL UNCERTAINTY ANALYSES "TENTATIVE SCHEDULE 35 L -- - - - - - - --

SCHEDULE IS BOTH AMBITIOUS AND OPTIMISTIC, AMBITIOUS BECAUSE OF EXTENSIVE NEW ANALYSES REQUIRED AND THE FACT THAT EVERY TASK IS ON CRITICAL PATH OPTIMISTIC BECAUSE IT ASSUMED:

o NO PROBLEMS WITH CODE PACKAGE LINKAGE o NO PROBLEMS WITH PLANT ACCESS FHYSICAL ACCESS INFORMATION ACCESS o NO ADDITIONAL SOURCE TERM RUNS NEEDED .

REBASELINING CONTAINMENT TREE PEER REVIEW o N0 NEED TO MAKE RISK-SIGNIFICANT CHANGES TO CONTAINMENT TREES OR TO REBASELINING EFFORT BASED ON RESULTS OF PEER REVIEWS s

. 36 q

REASONS FOR SLIPS

1. LEARNING CURVE

2. DIFFICULTIES IN INTEGRATING SOURCE TERM CODE PACKAGE (STCP)

3. ERRORS IN RUNNING STCP

4. NEED FOR ADDITIONAL STCP RUNS (SURRY)

5. NEED TO LOOK CLOSELY AT RECOVERY ACTIONS AND RISK REDUCTION MEASURES

6. ACCIDENT SEQUENCE ANALYSIS PEER REVIEW S

7. CONTAINMENT EVENT TREE PEER REVIEW

8. NEED TO CONSIDER ALTERNATE R0D INJECTION (ARI) AND AUGMENTED STANDBY LIQUID CONTROL SYSTEM (SLCS) FOR PEACH BOTTOM

9. SOME DELAY IN ACQUIRING PLANT INFORMATION FOR PEACH BOTTOM

10. NEED TO ADDRESS UNCERTAINTIES IN A MORE REALISTIC MANNER 37

OVERALL CONCLUSIONS RE: SCHEDULES o SLIPS ARE OCCURRING IN INDIVIDUAL SCHEDULES, AS ANTICIPATED o MAJOR STARTUP AND LEARNING CURVE PROBLEMS WERE IDENTIFIED EARLY AND LARGELY AVOIDED o THE ORIGINAL SCHEDULE FOR NUREG-1150 ANTICIPATED SOME STARTUP PROBLEMS AND REMAINS UNCHANGED, I.E.:

S DRAFT TO BE PUBLISHED FOR PUBLIC COMMENT MID-SUMMER 1986 l

l l

e 38

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l CONSIDERATION OF UNCERTAINTY 1

THERE ARE SIX BASIC AREAS THAT CONTRIBUTE TO UNCERTAINTY IN RISK PREDICTIONS:

o BASIC EVENT DATA (COMPONENT FAILURE RATES, HUMAN ERROR PROBABILITIES) o PRE-CORE-MELT PHENOMEN0 LOGICAL CONSIDERATIONS ASSOCIATED WITH ACCIDENT SEQUENCE FREQUENCIES (SUCCESS / FAILURE CRITERIA) o PHENOMEN0 LOGICAL CONSIDERATIONS PERTAINING T0 CONTAINMENT LOADS, FAILURE MODES, AND FAILURE LIKELI-f:

HOODS o PHENOMEN0 LOGICAL CONSIDERATIONS ASSOCIATED WITH ESTIMATES OF SOURCE TERMS IN THE CONTAINMENT ATMOSPHERE AS A FUNCTION OF TIME AND ACCIDENT SEQUENCES o METEOROLOGICAL VARIABILITY AND EFFECT OF VARIATIONS IN IMPORTANT CONSEQUENCE MODEL PARAMETERS o EFFECT OF EMERGENCY RESPONSE ASSUMPTIONS 3 C'

O' GENERAL PRINCIPLES FOR ANALYSIS OF RISK-DOMINANT UNCERTAINTIES PURPOSE IS TO FOCUS ATTENTION ON IMPORTANT ANALYTICAL ASSUMPTIONS AND ASSESS THEIR UNCERTAINTIES IN TERMS OF REASONABLE RANGES OF MEAN ESTIMATES OF CORE MELT FREQUENCY AND RISK c IDENTIFY 10-15 PARAMETERS OF MOST IMPORTANCE TO CORE MELT FREQUENCY AND RISK ESTIMATES o DETERMINE REASONABLE, CREDIBLE RANGE IN WHICH ACTUAL VALUES OF DRIVING PARAMETERS SHOULD LIKELY BE FOUND (ABOUT 90 PERCENT DEGREE OF BEllEF) 9 o CALCULATE RANGE OF MEAN VALUES BY COMBINING REASONABLE I RANGES OF MOST IMPORTANT PARAMETERS USING STATISTICAL SAMPLING APPROACH l

1 o RANGE OF MEAN VALUES WILL NOT BE EXPRESSED IN TERMS OF l ,

FORMAL STATISTICAL B0UNDS 3:o

' , :. ~.

ISSUE CLOSURE FOR SARRP/NUREG-1150 o INVESTIGATE RISK.IMPORTANT ISSUES INHERENT IN ASSESSING ACCIDENT PROGRESSION, CONTAINMENT BEHAVIOR, AND SOURCE TERMS o IDENTIFY DOMINANT PARAMETERS AND PERFORM SENSITIVITY STUDIES OVER A REASONABLE UNCERTAINTY RANGE TO ASCERTAIN IMPACTS ON RISK o DEVELOP NRC STAFF POSITION AND RATIONALE ON PARAMETER VALUES TO BE USED IN NUREG-1150 o TYPICAL TOPICS INCLUDE NON-VOLATILE (LANTHANIDE, i ACTINIDE) RELEASE FRACTIONS DURING CORE-CONCRETE INTERACTION, DIRECT HEATING, HYDR 0 GEN GENERATION AND OXIDATION WITH RESULTANT CONTAINMENT LOADING,

! REVOLATIZATION, RCS NATURAL CIRCULATION, BWR ATWS l

ACCIDENT ANALYSIS, BWR VENTING o OTHER ISSUES TO BE IDENTIFIED 1 -

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.PUBLIC COMMENTS ON NUREG-0956 o COMMENT PERIOD:

BEGAN AUGUST 7, 1985 EXTENDED FROM 90 TO 150 DAYS ENDS JANUARY 7, 1985 o LETTERS RECEIVED AS OF OCTOB$R 30, 1985:

EXTENSION OF COMMENT PERIOD 20 TECHNICAL 8 28 o REASONS CITED FOR EXTENDING COMMENT PERIOD:

TIME NEEDED TO GET NOTICE OF REPORT AND REPORT ITSELF MANY INTERESTED PEOPLE ON VACATION REPORT WILL HAVE SIGNIFICANT IMPACT ON REGULATIONS b

t.

REPORT IS LENGTHY AND HIGHLY TECHNICAL o TECHNICAL COMMENTS:

TECHNICAL COMMENTS HAVE JUST BEGUN TO ARRIVE EVALUATION OF TECHNICAL COMMENTS TO BEGIN S00N DISPOSITION OF COMMENTS TO BE DESCRIBED IN FINAL NUREG-0956 l

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D STATUS OF SOURCE TERM CODE PACKAGE o CODE PACKAGE NOW BEING USED IN SARRP o 0BSOLETE MODEL OPTIONS BEING REMOVED IN PREPARATION FOR PUBLIC RELEASE o RELEASE OF DOCUMENTATION AND TAPES EXPECTED IN DECEMBER o MANUAL FOR CODE PACKAGE WILL RELY ON MANUALS FOR INDIVIDUAL CODES I.

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3.! NN~R STAFF PRESENTATION TO THE '

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SUBJECT:

POTENTIALLICENSINGUSESOFSOURCElERt1RESEARCH DATE: AUGUST 2,1985 PRESENTER: L. G. (IRRY) HUU%'l PRESENTER'S TITLE / BRANCH /DIV:' CHIEF, ACCIDENT EVALUATION BRNCH DIVISION OF SYSTElG INTEGRATION, tRR PRESENTER'S NRC TEL. NO.: 492-7880 SUBCOMMITTEE: SE\ERE ACC10Elli

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10 GENERAL AREAS OF REGULATORY OVERSIGHT DEPENDENT UPON ACCIDENT SOURCE TERMS e

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BACKGROUND & CURRENT PRACTICE e

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PR5LIMINARY COST / BENEFIT ASSESSENT -

SCHEDULE FOR I W LB elTATION PROPOSED . -

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SEARCH F,0R RISK OUTLIERS TO USE SOURCE TERM RESEAR

, TO IMPLBENT SEVERE ACCIDENT POLICY 0 IDCOR WT110DOLOGY DEVELOPENT 0 NRC REVIEW 5

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.,T.. 0 INDUSTRY APPLICATION

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CONTAltfENT PERFORMANCE

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0 TID-14844 LCfA DOSE ASSESSENT WITHOUT KNOWN f-'

PRODUCT ATIENUATION KCHANISMS ASSLNING R NOBLE GASES 8 10 DINE O CONTAlffENT LEAKAGE EMPHASIZED, NOT CONTAltKNT ULTIM STRENGTH 0 INSTANTANE0LjS ELEENTAL 10 DINE ASSLNPTIONS FROM TID CCUTROLS CURRENT PRACTICE, O DBA LOCA THAT RESULTS IN FEW OR NO FUEL F

- ESTABLISH DESIGN TEMPERATURE & PRESSURE, AEROSOLS NOT

CONSIDERED t

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,. POTENTIAL CHANGES - SHORT TERM

• 0 REASSESS BASES FOR LOW LEAK RATES & T h:c.-

u WITN EMPHASIS ON CONTAlttENT INTEGRITY a .

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'i 1- 0 CONSIDER ADMINISTRATIVE CONTROLS OR PENE U GUIDELINES TO IBPROVE DETECTION OF CO O CONSIDER TESTING REQUIREENTS FOR CONT 0 MDDIFY CLOSURE TIE CRITERIA FOR VALVES CONTAlwENT

• BASES FOR CHANGES WILL REQUIRE

1) ECHANISTIC ANALYSES OF FISSION PROD

2) DETERMINATION OF LEAKAGE PATHS IMPO

3) EXPLORE VIABILITY OF f0NITORING INTEGRIT t

4) DEVELOP BASES FOR TECH SPEC REVISIONS

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AREA 3 ENVIR0tfENTAL QUAllFICATION OF EQUIPENT

.Yj BACKGROUND & PRESENT PRACTICE t

l;, 0 TID-14844 BASED SUBSTANTIAL CORE-ELT RADIATION LEVELS O TEMPERATURE & PRESSURE CONDITIONS CONSISTE LOCA PRODUCING ONLY A FEW % FUEL FAILURES O TMI INFORMATION BEING EVALUATED 0 SCE BEYOND DBA SURVIVABILITY BEING IMPLB' of H2RULE POTBRIAL CHANGES

' - 0 UNKfGN, BUT TO BE STUDIED IN TERMS OF:

1) EQUlftB R FUNCTION

2) POTENTIAL MARGINS P

3) RELATIVE CONTRIBUTIONS TO RISK

4) TMI INFORMATION O 5) CHEMICAL 8 PHYSICAL FORMS AND QUANTITIE I PRODUCTS AND t0N RAD AEROSOLS L .

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EXPOSURE EPZ  !

O FEMA EVALUAES STATE AND LOCAL PLANS A POTENTIAL CHANGES .

0 CORRECT MISCONCEPTION THAT EVACUATIO TO 10 MILES OR FURTHER -

! 0 USE CONCEPT OF GRADED RESPONSE THAT R VARIATIONS WITH DISTANCE & TIE Q.

O CONSIDER RISK VARIATIONS OF REACTOR T .t n,

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'!;l PROPOSED ACTIONS ii.:

'3 0 DRAFT REVISIONS TO 10 CFR 50 BASED UPON NEW SOURCE TERM INFORMATION CONSIDERING 3 ALTERNATIVES 7

? 1) PLANT CLASSIFICATION BASED UPON APPARENT RISK PROFILE l..':

4 2) SITE SPECIFIC EERGENCY PLANS

3) GRADED RESPONSE O REVISE GUIDANCE DOCitENTS (I.E., FEMA /NRC NUREG-0654, NUREG-1082)

O COORDINATE WITH EPA ON POTENTIAL CONFLICTS WITH PROTECTIVE A GUIDES l

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- AREA 5 12 ACCIDENT CONSEQUENCES & INDBHIFICATION

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0 ONSITE PROPERTY DAMAGE INSURANCE REGULATED THROUGH g: 10 CFR 50,5f4(W)

., 0 0FFSITE INDStiFICATION UNDER PRICE / ANDERSON t

0 E N,0, DETERMINATION CRITERIA (10 CFR ll40) SET WELL BELOW CONSEQUENCES OF MANY SEVERE ACCIDENT RELEASES 0 NEPA EVALUATIONS OF ACCIDENTS POTENTIAL CHANGES 0 DETERMINE ItPACT OF RESEARCH ON PROPOSED ATNDE NTS TO ONSITE REQUIREN NTS, REVIEW NUREG/CR-2601, AND PROPOSE .

RULE CHANGE IF NECESSARY

O CONSIDER ADVISING CONGRESS OF CHANGES IN OFFSITE RISK ri PERCEPTION IF WARRANTED m,

1 3 0 USE fEW PETHODOLOGY IN FLTTURE NEPA REVIEWS (STARTING WITH S0lffH TEXAS OL REVIEW)

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'i 0 INSTANTANEDUS FISSION PRODUCT RELEASE ASSLPED, 5

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0 ELIMINATE NEED FOR AUTOMATIC INJECTION OF CAUSTICS IN O CONTAlffENT SPRAYS AT PWRS, B(JT CONTINUE TO REQUIRE l CONTROL OVER Sl@ PH CREDIT BWR CONTAlffENT SPRAYS AND SUPPRESSION POOLS 0

O DO NOT BACWIT ELEENTAL IODINE FILTER GUIDANCE a DEVELOP REPLACEENT GUIDANCE FOR ICDINE, OTHER FISSION PRODUCTS &

AEROSOLS IN BOTH ROUTINE AND ACCIDENTAL RELEASES.

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kk$ 4 ff; PRESENT PRACTICE V.'v l 0 MONITORS FOR TID-14844 RADIATION LEVELS OR GREA GUIDE 1,97) '

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POTENTIAL CHANGES O REASSESS INSTRlPENTATION REQUIRE 0F ACCIDENTS O REASSESS OFFSITE INSTRitENTATION NE I

IN TIMING & FISSION PRODUCT MIX TO DETE ,

1 0F 10 CFR 140 >

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.: 0 INCLUDES HEALTH EFFECTS, PROPERTY DAMAGE, CR0P INTE

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!"! EVACUATION AND DEC0KTAMINAT10N COSTS, a USES OF K1

. 0 USED TO ASSESS PREVENTION & MITIGATION FEATUR O INFLUENCES PUBLIC OPINION POTENTIAL CHANGES

.. O HAS INFLUENCED POLICY ON USE OF KI 0 POTENTIALLY LOWER LEVELS OF 0FFSITE CONTAMINATION,

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PLANTS LIKE SURRY, BUT WE DON'T KNOW ABOUT OEERS

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8 0 REASSESS DESIGN BASIS ACCIDENT ETHODOLOGY, w;

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2) MODEST CHANGES TO TID-14844 RELEASE ASSLNPT E.

.y 3) RESTATE DBA A) DETERMINISTICALLY USING THE PRESENT TY ACCEPTANCE APPROACH i

, B) ESTABLISH ESF PERF0WANCE CRITERIA FOR ES

4) USE OF RISK BASED APPROACH O BACKFITS ON EXISTING PLANTS, OR ADDED SAFETY FEAT J

NEW DESIGNS, LIKELY TO BE LESS COST-EFFECTIVE, a

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SAFETY EVALUATIONS hi! i t$

M 1)Q 81 PRESEN'T PRACTICE u:

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X- 0 FOR SCREENING NEW SAFETY ISSUES STAFF USES WASH-14 N SOURCE TERMS IN ASSESSENTS OF

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1) RISK IMPORTANCE AS INDICATED BY 50 MILE POPULATIO

2) VALUE/ IMPACT OR RATIO OF RISK REDUCTION TO TOTAL C (INDUSTRY AND NRC) 0 FOR ASSESSING BACKFITS USE MORE. DETAIL IN VALUE/IFP ASSESSMENTS, BUT IN SOT CASES USING NEW SOURCE TERM INF0 l

- 0 LICENSING REVIEWS, INCLUDING ABENDENTS, OFTEN USE

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TlD-14844 ASSTFTIONS OF NOT ONLY FISSION PRODUCT b?$ RELEASES, BUT OT!ii.R ASS &PTIONS OF TRANSPORT, a> 9, ,

.'1 : DEPOSITION & BIOLOGICAL UPTAE

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@y. PRESENT PRACTICE & BACKGROUND Vi O TlD-14844 WITH LOCA, SUBSTANTIAL CORE MELT, AND NO CONTAI E NT 5' FAILURE COUPLES SITING & PLANT DESIGN THRU EVALUATIONS O Q. .

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1) EXCLUSION AREA BOUNDARY m  ;

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2) LOW POPULATION ZONE

3) PERFORMANCE CRITERIA FOR CERTAIN ESFS

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O CORE E LTS WITH CONTAI E NT FAILURE INFLUENCED REFDTE SITING, SUPPORTED BY WASH-1400

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l$_. 0 1979 RECC41ENDAT10N TO DECOUPLE SITING & DESIGN (NUR

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l: /' - THROUGH RULB%K!NG

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q 7; 1 EVALUATIONS OF SOURCE TERM 3 & SAFETY G0ALS

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p POTENTIAL CHANGES  :

1. DBA 0 DEVELOP NEW SITING DBAs a PERFORWICE CRITERIA FOR ESFs i

0 ELIMINATE DBAS FOR SITING EVALUATIONS 8 REPLACE WITH REQUIREENTS FOR MINIM COMPLIENT OF ESFS WITH SPECIFIC PERFORMANCE CRITERIA 8 A MINI M SET OF SITE CHARACTERISTICS 1

2. BEYOND DBA 0 RECONSIDER RBiDTE SITING POLICY USING NEW SOURCE TERNS INFO THRU RISK REBASELINING

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