ML19351D903

Discusses Current Accident Evaluation Practices in Siting & Licensing of Nuclear Power Plants.Lists Results of Current Accident Evaluation Practices.Accident Analysis, Site Evaluation & Other Related Info Encl

ML19351D903
Person / Time
Issue date:	02/22/1978
From:	Minogue R NRC OFFICE OF STANDARDS DEVELOPMENT
To:
References
SECY-78-111, NUDOCS 8011200358
Download: ML19351D903 (49)

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February 22, 1978 uNitta starts NUCLEAR REGULATORY COMMIS$10..

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20985 INFORMATION REPORT For:

The Commissioners From:

Robert B. Minogue, Director, Office of Standards Development bh Thru:

Executive Directer for Operations subject:

CURRENT ACCIDENT EVALUATION PRACTICES IN SITING AND LICENSING OF NUCLEAR POWER PLANTS

Purpose:

To inform the Commission of current accident evaluation practices in the siting and licensing of nuclear power plants.

Subsequent papers will discuss alternative accident evaluation practices, revisions to current regulatory requirements and forward looking policy in this area.

Backgrouno:

The Office of Standards Development (50) staff briefed the Com-M mission on the results of an NRC staff review on reactor site I#

evaluation policy (SECY-76-286) and the developing plan for tQ revising NRC nuclear facility siting policy and practices (SECY-75-286A).

o Following the briefing on January 13, 1977, a memorandum from the Secretary of the Commission, dated Janucry 27, 1977, directed the EDO to prepare four different policy statements on major reactor siting issues, one of which was to address nuclear reactor accident evaluation practices for promulgation by the Commission.

o Based on the June 17, 1977 briefing on the general policy statement on nuclear reactor site evaluations (SECY-77-288),

SD was directed by the Secretary's memorandum of June 30, 1977 to address only current siting policy and practice.

Based on these instructions, this paper was prepared by SD in a joint effort with NRR to reflect only current accident evaluation practices in the siting and licensing of nuclear power plants.

Discussion:

Our response to the memoranda from the Secretary will consist of the following papers:

Contact:

F. Anderson, 50 443-5317 H

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B012"003 1

O The Commissioners 2

1.

This information paper which describes current accident evaluation practices in the siting and licensing of nuclear power plants.

It will serve as the point of departure for subsequent papers discussing Commission policy in this area.

2.

A policy paper which discusses possible changes in our current accident evaluation practices.

A discussion of major concerns within this area and recommendations for resolving the issues and modifying the regulation will be included.

This paper should be completed by July 1978.

3.

A' policy paper requesting the Commission to publish revised regulations for public comment with a statement of policy to be used for the statement of considerations for the changes.

This paper would be completed following the input from the Cou ission on the previous paper.

A discussion of current accident analysis and site evaluations is attached as Enclosure "A".

The purpose of this enclosure is to describe and summarize the role of accident analyses in the evaluation of proposed sites for current nuclear power reactors.

The reviews of nuclear power reactors to protect the health and safety of the public are performed to implement the requirements of the' Atomic Energy Act of 1954, as amended.

Accident evalua-tions for system design in nuclear power reactors are performed to meet the needs and requirements of 10 CFR Part 50 for licens-ing such facilities.

Accident evaluations for site suitability purposes are performed to meet the needs and requirements of 10 CFR Part 100 for siting nuclear power reactors.

The accident evaluations performed to determine the adequacy of system design's for plant safety considerations may not and need not be the same -

as those performed to determine the suitability of a selected site for the location of a nuclear power reactor.

For example, early site review considers the suitability of a site independent of specific facility design.

Nonetheless, bounding envelopes for plant and site are used in these separate reviews because ulti-mately the staff must determine whether a specific plant at a specific site provides reasonable assurance that the health and safety of the public will not be endangered by operation of the nuclear facility in the proposed manner.

This determination is based upon the combined results of the two accident evaluation reviews performed by the NRC staff.

A discussion of current regulatory requirements for accident evaluation in the siting and licensing of nuclear power plants is

The Commissioners 3

attached as Enclosure "B".

In summary, the regulations contain cwo principal aspects for accident evaluation.

One aspect relates to site features and the other aspect to the design features of engineered safety features in the nuclear facility.

The engi-neered safety features such as the emergency core cooling systems and reactivity control systems may be designed to prevent or mitigate a postulated safety system failure.

Other engineered safety features such as containment systems (controlled leakage) and combination spray / filter systems (atmospheric cleanup) may be designed to mitigate the radiological consequences from such postulated reactor accidents.

The site features may be con-trolled by site selection to minimize the radiological conse-quences to the public from releases of radioactive material resulting from postulated reactor accidents.

These site features include such physical characteristics as meteorology for airborne dispersion and hydrology for waterborne dispersion and such 4

features as land use for pathway concerns and population distri-bution for magnitude of individual and integrated radiation exposures.

Other important site features include distances to nearest site boundary, low population zone and population centers.

A general comparison of the facility design features and the consequence analysis parameters has been made for the facility /

site combinations approved for licensing over the past 20 years.

The result of this comparison is attached as Enclosure "C".

The general comparison for accident evaluation purposes has used some of the important features and elements of the facility design from the plants reviewed and licensed up to the time of issuance of 10 CFR Part 100 and the associated technical document, TID-14844, and similar features and elements from the plants currently being reviewed and licensed under the same regulation with the Standard Review Plan and Regulatory Guides.

Evolution of plant design has forced changes in the analysis parameters for the selected accidents used in the NRC staff accident evaluations.

As the plants have become larger, consequences of a major accident have grown.

Correspondingly, engineered safety features have been added to greatly reduce the likelihuod of such consequences from major accident.

Hence, the emphasis in the application of Part 100 has shifted toward evaluation of mitigative measures.

However, the overall balance of site and facility required by Part 100 has not changed.

These resulting changes in conse-quences and risks are discussed in the enclosure.

The results of current accident evaluation practices are as follows:

1.

The available array of engineered safety features is suffi-cient to justify a very small exclusion area and short LPZ

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The Commissioners 4

boundary distance.

Nonetheless, the precise balancing between site isolation (size of exclusion area and LPZ) and reliance on engineered safety features is largely discre-tionary.

Furthermore, for sites with high population densi-ties, additional criteria come into play to balance the proposed site against alternatives.

2.

From the accident analysis standpoint, the closest distance to a center of population is regulated only with respect to determining that it must be at least 1-1/3 times the LPZ distance.

However, a comparison of computed radiological cor. sequences against the reference guideline dose values in 10 CFR Part 100 is not, by itself, a sufficient basis for determining whether the site isolation distances are ade-quate for locating a proposed nuclear facility.

3.

In keeping with the purpose of -10 CFR Part 100, site suit-ability ultimately is a judgment made by the staff based upon consideration of all the factors and criteria con-tained in that part.

Coordination:

The Director of the Office of Nuclear. Reactor Regulation concurs in this paper.

The Director of the Office of the Executive Legal Director has no legal objections to this paper.

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Robert B. Minogue, Director Office of Standards Development

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Enclosures:

"A" - Accident Analysis and Site Evaluation "B" - Discussion of Current Regulatory Requirements for Accident Evaluation Ir Siting and Licensing of Nuclear Power Plants "C" - Comparison of Elements in Accident Evaluation Practice for Light-Water-Cooled Reactors DISTRIBUTION Comissioners Comission Staff Offices Exec Dir for Operations ACRS Secretariat 1

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ENCLOSURE A ACCIDENT ANALYSIS AND SITE EVALUATION

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e ENCLOSURE A ACCIDENT ANALYSIS AND SITE EVALUATION The purpose of this discussion is to describe and summarize the role of accident analyses in the evaluation of proposed sites.

The history of nuclear power reactor site evaluations has been included to show the development of staff practices. An adequate understanding of the current practices requires some discussion and clarification of the bases for the criteria in 10 CFR Part 100.

Where necessary, this discussion describes the development of the Part 100 criteria and thcir bases.

I.

Safety Reviews Under the Atomic Eneroy Act Since 1955, 10 CFR Part 50 has served as the heart of the Commission's safety regulations governing nuclear power reactor design.I Site-related considerations have been an integral part of the licensing reviews; from the outset, there has been a recognition of the need to consider the risks associated with possible accidents as an important element in making a IAEC 23/22 (March 30, 1955) submitted 10 CFR Part 50 for Commission approval.

Contaf:.2d in that first version of Part 50 was a requirement that an appli-cation contain a "... description...in sufficient detail to allow an evalua-tion of the adequacy of the various means proposed to minimize the proba-bility of danger from radioactivity to persons both on and off-site".

1 Enclosure "A"

o determination of the suitability of a proposed site for locating a nuclear facility.2 In late 1958, the Comission's staff began efforts to develop specific siting criteria to serve as aids and guides to be used in the selection and evaluation of sites. These efforts culminated in April 19G2 with the issuance of 10 CFR Part 100.

The criteria in Part 100 did not decouple the reviews of facility design from those of the site but rather (as will be discussed later) provided a framework for judging whether a particular facility / site combination was acceptable.

Part 50 still contains pro-visions regarding the evaluation of the proposed site.

For example, one of the findings required by the Commission in granting a construction permit is that "...there is reasonable assurance that...taking into con-1 sideration the site criteria centained in Part 100 of this chapter, the proposed facility can be constructed and operated at the proposed location without undue risk to the health and safety of the public."3 Part 50 2AEC 646/45 (March 15,1956) contains correspondence from W. F. Libby, Acting Chairman, to Senator Hickenlooper which noted "there is always present, regardless of the remoteness of its probability, a finite probability of the occurrence of an event, or series of events, the result of which is the release of unsafe quantities of radioactivity to the surrounding area.

...it is therefore more desirable from the standpoint of safety alone to locate reactors in areas of low, rather than normal, population density." That same letter also describes the approach now tenned " defense-in-depth" and notes that protection of the public relies on both provisions in design and relative site isolation.

310 CFR Part 50,' Section 50.35(a).

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Enclosure "A" l

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also requires that an application for a construction permit include "a description and safety assessment of the site on which the facility is to be located....

Such assessment shall contain an analysis and evalua-tion of the major structures, systems and components of the facility which bear significantly on the acceptability of the site under the site evalua-tion factors identified.in Part 100...."4 (emphasis added)

The basic thrust of the facility / site evaluation process was summarized 5

in the Statement of Considerations for the proposed 10 CFR Part 100 as follows:

" Judgment of suitability of a reactor site for a nuclear plant is a complex task.

In addition to nomal factors considered for any industrial activity, the possibility of release of radio-active effluents requires that special attention be paid to physical characteristics of the site, which may cause an inci-dent or be of significant importance in increasing or decreasing the hazard resulting from an incident.

Moreover, the inherent characteristics and the specifically designed safeguard features of the reactor are of paramount importance in reducing the possibility and consequences of accidents which might result in the release of radioactive materials.

All of these features of the reactor plus its purpose and method of operation must be considered in determining whether location of a proposed reactor at any specific site would create an undue hazard to the health and safety of the public.

" Recognizing that it is not possible at the present time to define site criteria with sufficient definiteness to eliminate the exercise of agency judgment, the proposed guides set forth below are designed primarily to identify a number of factors considered by the Conunission and the general criteria which are utilized as guides in evaluating proposed sites."

4 10 CFR Part 50, Section 50.34(a)(1).

5 Reactor Site Criteria, Notice of Propo;ed Guides, 26 FR 1224, February 11, 1961.

3 Enclosure "A"

The site evaluation factors specifically identified in Part 100 include the characteristics of reactor design and proposed operation (including power level, the extent to which generally accepted engineering standards have been applied to the reactor design, the extent to which the reactor incorporates unique factors having a significant bearing on the probability or consequences of an accidental release of mate? al, and the safety l

features that are engineered into the facility), population density and land use characteristics of the site, and physical characteristics of the site.

Where unfavorable physical characteristics of the site exist, the extent to which th!. design of the facility includes adequate compensating engineered safety features is an additional factor to be considered.6 To perform an assessment of a site that takes into account the above factors the staff evaluates the spectrum of possible accidents.

The spectrum of accidents is very broad, ranging from high probability events whose consequences are trivial, to very low probability events whose consequences may result in loss of life.

The potential for a serious accident is made acceptably low by design or administrative procedures, and, as warranted, by the provision of engineered safety features to mitigate the accident and/or the radiological consequences.

While as noted above, both Part 50 and Part 100 expressly provide for an assessment of the facility / site combination, current staff review 610 CFR Part 100, Section 100.10.

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Enclosure "A" i

s practice depends on a substantial decoupling of site characteristics from facility design reviews.7 That is, the majority of applications include a facility design that is the same as or very similar to one which has been reviewed in the past.

In such instances the balancing between facility design (including the degree of reliance on engineer-ing safety features) and relative site isolation can largely be accom-plished by reference to the earlier reviews, with any necessary adjust-ments to reflect site related differences in physical characteristics.

Similarly, our past experience and guidelines have established a general pattern of what are suitable characteristics for a site.

Regulatory Guide 4.7 recommends general site suitability criteria that have generally been found acceptable to the staff based on previous reviews.

II.

Part 100 Accident Evaluations Under Part 100, the quantitative analysis of the consequences of one postulated accident is required as a key test of facility / site suita-bility.

That accident is a " major accident, hypothesized for purposes of site analysis or postulated from consideration of possible accidental 7Early site reviews, for example, are possible only if the basic premise is made that the facility design will be substantially the same as one previously reviewed by the Comission.

Standardized plant designs typically specify a site characteristic envelope within which no spe-cial considerations regarding site suitability need be reviewed.

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Enclosure "A"

events, that would result in potential hazards not exceeded by those from any accident considered credible".8 Part 100 does not provide either criteria or guidelines on the various assumptions that must be made to translate the assumed fission product release into estimated consequences.

That is, there are no statements 9

in Part 100 regarding the magnitude or assumed time rate of release of the fission product source term, the transport of the radioactivity within the facility, or the assumed effectiveness of engineered safety features. The meteorological condition assumed to exist at the time of the accident was to be pertinent to the site while the expected leak rate from the containment was to be demonstrable.10 However, Part 100 810 CFR Part 100, footnote 1.

See AEC-R 2/19 for a discussion of this event.

A typical definition of credible included any event resulting from a single failure of equipment or one operational errcr that could cause a fission product-releasing accident.

The maximum credible acci-dent (MCA), at the time of writing of Part 100, was typically a loss-of-coolant accident which could result in some core melt with the contain-ment barrier functional but pressurized to that value resulting from the large pipe rupture (for plants of the type considered in the 1950's, this was regarded as a consistent set of assumptions; no credit was given to the possible effectiveness of " active" engineered safety fea-tures such as emergency cooling systems or containment sprays, but no special requirements were placed on such systems either).

For a further discussion of the MCA, see pp. 8-12 of TID-14844 or the paper by Clifford K.

Beck, " Safety Factors to be Considered in Reactor Siting" (June 1955).

90ther than the declaration footnote in 10 CFR Part 100 that states "such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release of fission products."

1010 CFR Part 100, Section 100.11(a).

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l refers to a procedural method and a sample calculation (TID-14844 March 23,1962) that can be used.

TID-14844 does include values for each of these various parameters.II There are six principal considerations involved in analyzing the con-sequences of the accident selected for purposes of site evaluation,12 each of which are discussed in Part 100 and in TID-14844.

They are:

1.

The maximum inventory of fission products likely to be present, usually related directly tr-

e power level.

4 2.

The quantity of fission products released from the primary system in the event of the maximum accident considered to have a credible likelihood of occurrence.

U In TID-14844 only the " passive" containment was considered in the analysis of the consequences of the maximum credible accident. This assumption was used in the reviews of applications prior to issuance

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of 10 CFR Part 100.

Subsequent to issuance of Part 100, credit was given for certain " active" engineered safety features, thereby per-mitting location of power reactors on smaller exclusion areas and closer to cities than the distances based on TID-14844.

For a dis-cussion of. the implications of greater reliance on active systems, see

" Engineering out the Distance Factor," a paper by Clifford K. Beck (September 25, 1963; AEC Press Release No. 5-28-63).

12Testimony of Robert Lowenstein before the JCAE, June 12, 1961.

7 Enclosure "A"

3.

The effectiveness of various engineered safety features in con-trolling the release of fission products to the environment.13 4.

The pattern by which the radioactive material released to the J

environment would be transported and dispersed to surrounding i

4 areas.

5.

The radiation doses to individuals that may result from exposure 4

to or uptake of the dispersed radioactive material.

6.

The radiation doses which are considered acceptable for such unlikely accident situations.

Since TID-14844 was issued, there have been many changes in power reactor designs, as indicated in Enclosure "C", and research and development has led to further improvements in engineered safety features, both factors of which have a bearing on site suitability.

Further, considerable experience has been gained in the review and analysis of current nuclear j

power reactor designs.

This base of experience has in turn led to the development of detailed staff guidelines for the design and analysis of such reactors. As a result, the evaluation model and assumptions stated 13As noted in 10 CFR 100, Section 100.11(a), the " expected demonstrable leak rate from the containment" is to be used in the consequence analysis, i

The assumed effectivenesslof the containment or any " active" engi-neered safety features is determined as part of the design reviews of Part 50; technical specifications resulting from those reviews are used as input to the Part 100 consequence assessments.

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Enclosure "A"

in TID-14844 are not used by the staff and have been replaced by the evaluation models and assumptions given in various Regulatory Guides and the staff's Standard Review Plan.

These differences are discussed fur-ther in Enclosure "C".

These documents provide specific guidelines suggested for use in each area of the calculation of accident conse-quences, from source term to dose conversion factors.

III. Site Suitability Calculations The accident conditions recommended for use in Part 100 reviews of current light-water-cooled reactors (LWR) remains numerically about the same as those in TID-14844.

Namely, the assumed fission product release is very large, of a magnitude generally associated with melted fuel, the centain-ment is assumed to remain functional, with the containment pressurized to a value corresponding to that associated with a loss-of-coolant accident (LOCA) from a major pipe break.I4 Based on this postulated source term in the containment volume (with the core inventory based on the ultimate or stretch power) and assumptions regarding the effectiveness of various engineered safety features which are provided to mitigate the consequences of the LOCA, an estimate is made of the fission products released to the atmosphere.

An assessment of the consequences of this release is made considering three different groups of population (according to areas surrounding the 14Regulatory Guides 1.3 and 1.4.

Given the current criteria in Appen-dix K to 10 CFR Part 50, the large source term can no longer be mechanistically tied to the loss-of-coolant accident.

9 Enclosure "A"

r plact).

The first is the exclusion area, within which people are con-sidered to be very mobile.

Where transients or residents might exist within this area, the computed deses must be within the guideline values of Part 100 during the period of evacuation.15 The dose that would be received by an individual at any point on the exclusion area boundary for two hours immediately following onset of the postulated fission product release is also evaluated.

This calculated dose should be equal to or less than the stated guideline dose levels of 300 rem to the thyroid or 25 rem t'o the whole body for the size of exclusion area to be suitable for the proposed location of the nuclear facility.IO The second group are those individuals within the area termed the low population zone (LPZ) in Part 100.

This group is not under'the control of the applicant, but there must be reasonable assurance that protective measures such as evacuation could be taken on their behalf.

The dose that could be received by an individual at the outer boundary of the LP?

who is exposed to the radioactive cloud (during the entire period of its

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passage) resulting from the postulated fission product releases is evaluated.

This calculated dose should be equal to or less than the stated guideline dose levels (300 rem thyroid or 25 rem whole body) for the size of the LPZ to be suitable for the proposed location of the nuclear facility.16 15This is a staff practice and is not a part of the regulation.

16Regulatory Guides 1.3 and 1.4 provide a suitinary of the assumptions to be used for this calculation.

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10 Enclosure "A"

The third group is represented by the population center and individuals in this group are generally considered to be less mobile than those within the LPZ.

Since specific dose guidelines are not stated for this group, dose calculations are not required to determine the exposures to this group or site suitability.

However, the population center distance is required to be at least 1-1/3 times the distance from the reactor to the outer boundary of the LPZ.

The criteria in Part 100 state that where very large cities are involved, a greater distance may be r.ec-essary because of total integrated population dose considerations.

These calculated doses are based on exposure to airborne radioactivity; calculations of possible doses from other pathways (such as ingestion of contaminated milk following deposition of iodine on land) are not normally performed.17 Given that the assumed fission product release and the exposure criteria (dose guidelines) are fixed, the suitability of a site from the stand-point of calculated accident consequences depends largely on the assumed I7The rationale for this practice ste'as directly from two related factors:

(1) the relative ease of preventing intake of highly contaminated food or water and (2) placing relatively greater emphasis in the siting decision on preventing acute fatalities.

One of the principal criteria of Part 100 is the guideline dose levels.

As stated in Note 2 to Part 100, these guideline dose levels were set forth as reference values, which can be used in the evaluation of reactor sites with respect to potential reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation.

However, in the actual event of a release, decisions on emergency countermeasures would be made with the objective of minimizing doses.

This intended applicability of the guide-line dose levels is discussed in AEC-R 2/25 (2/8/61).

11 Enclosure "A"

effectiveness of each of the various engineered safety features provided to mitigate the consequences of the postulated LOCA.

As noted in the TID-14844, the containment was the only engineered safety feature con-sidered in those dose calculations.

Currently, a variety of other fea-tures are relied upon to mitigate the potential consequences of the postuhted LOCA and thus " engineer out the distance factor" and reduce the calculated exclusion area boundary and low population zone distances from those which would result from usa of the TID-14844 evaluation model and assumptions.

Current designs which employ dual containments and iodine removal systems such as sprays and filters have the capability for removing virtually all of the iodine assumed to be released in the postulated LOCA.

By employing such designs, it is possible today for nuclear power plants to be located at sites with a very small exclusion area and a short low population zone distance and still meet the dose criteria of Part 100.

For example, the Bailly site has a minimum exclusion area distance of 188 meters (0.12 mile), which is comparable to the TID-14844 distance of 1 mile for the proposed power level.

The Midland site has a low popula-tion zone radius of 1600 meters (one mile), which is comparable to the TID-14844 distance of 19 miles for the proposed power level.

The major change from the TID-14844 evaluation has to do with the assumed effectiveness of various " active" engineered safety features to mitigate 12 Enclosure "A" r

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o the cons.equence of accidents. Although there have been numerous changes in the nadels and assumptions used to calculate meteorological dispersion i

and radiological dose, the principal accident assumptions in the calcula-

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tion remain the same as given in the TID-14844 It should be noted that the postulated LOCA may no longer be the limiting accident for those plant designs which employ a full complement of engi-1 neered safety features to mitigate the consequences of the LOCA.

Those plants may have other controlling design basis accidents, such as those involving steam generator tube ruptures or steam line breaks outside o' containment, whose conseq;ences are not affected significantly by the containment-related engineered safety features provided to accommodate the postulated LOCA.

Therefore, in evaluating the suitability of a proposed site for a reactor plant, the staff either confirms that the radiological consequences of the LOCA with a coincident large source term is limiting, or that there is some other limiting design basis accident, and, lastly, that none have consequences which exceed the guide-line dose values of Part 100.

In summary, an important part of the process of determining the suita-bility of a site involves comparison of the calculated radiological consequences of the postulated LOCA (or some other more limiting event) at the proposed exclusion area boundary and low population zone distances, against the guideline dose levels given in 10 CFR Part 100.

However, it 13 Enclosure "A"

l should be emphasized that this procedure was devised primarily as an aid in the evaluation of sites from the standpoint of the various evaluation factors identified in 10 CFR Part 100.

The suitability of a site for any particular reactor cannot be determined from the criteria in part 100 solely on the basis of calculations of the radiological consequences of one assumed release at two selected distances.

Application of the reactor site criteria depends very substantially upon the evaluation of the specific design features of the particular reactor involved and the specific characteristics of the selected site, taking into account all of the evaluation factors identified in Part 100.

IV.

Consequence Related Factors Not Explicitly Addressed by Calculations To help illustrate the above point, it may be useful to discuss several reasons why the numerical consequence results of the Part 100 accident analyses must be tempered by other considerations.

As stated in the Statement of Considerations to 10 CFR Part 100,18 "These guides...are intended to reflect past practice and current policy of the Commission of keeping stationary power and test reactors away from densely populated centers." The Statement of Considerations also noted that "...since acci-dents of greater potential hazard than those commonly postulated as 'repre-senting an upper limit are conceivable, although highly improbable, it was considered desirable to provide for protection against excessive exposures to people in large centers....

The population center distance 1827 FR 3509, April 12, 1962.

14 Enclosure "A"

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was added as a site requirement when it was found for several projects evaluated that the specification of such a distance would approximately fulfill the desired objective and reflect a more accurate guide to current siting practices."I9 l

As discussed earlier, the reactor site criteria in Part 100 state only that the distance to the marest boundary of a densely populated center containing more than about 25,000 residents is at least 1-1/3 times the distance to the outer boundary of the low population zone.

A currently i

designed LWR could, with a full complement of engineered safety features, meet the Part 100 dose guidelines even with a short low population zone distance (as in the one mile LPZ at Midland).

Using only the rule that the population center distance must be at least 1-1/3 times the LPZ distance, it would be possible to locate the reactor close to a city or a densely populated area.

Before such a situation would be found accept-able, it would be necessary to determine that the aforementioned policy objective regarding metrop611 tan siting was met.

l9That the intent of the population center distance was based on what are currently called Class 9 accidents (and specifically core melt with failure of containment) is clear from the Statement of Considera-tions to the 1961 version (AEC-R 2/25; AEC-R 2/19 provides a short sumary of the background leading to the Part 100 reactor siting cri-teria, including the role of Class 9 accidents in siting at that time).

Based on the assumptions and reactor design features in use at that time, the consequences of core melt with containment failure were not likely to result in acute fatalities at population center distances 1-1/3 times the LPZ distance (as detennined from the same event with containment -intact).

15 Enclosure "A"

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With regard to nearness to cities, the numerical calculations of the LPZ (and hence population center distance) are based on " credible" or design basis events, and the LPZ distance, by these calculations, can be made short if enough features to mitigate the consequences of design basis events are provided. One objective in requiring the minimum population center distance is to provide some measure of protection against yet more severe events.

The suitability of the site with regard to nearness of large populations must be considered and a judgment made as to whether the computed population center distance fulfills the objective of keeping power reactors away from densely populated centers.

1 More specifically, the population center distance was included as a cri-terion to satisfy two objectives, one of which was to minimize cumulative exposure from any release and another of which was to control the conse-quences of events beyond the design basis.

With the practice of including credit for active engineered safety features, the second objective is no longer directly implemented by the results from the numerical calculations of the consequences for the postulated LOCA.

Secondly, none of the criteria directly control the size or distribution of the population (or population density) in the vicinity of the facility; the only stated criteria are qualitative in nature.

Consequently, it has been necessary ror the staff to develop additior.al guidelines on population density and use character-istics of the site environs to supplement the Part 100 accident analyses 16 Enclosure "A" 1

and further the implementation of the policy objectives of the reactor site criteria.20 In any event, the application of reactor site criteria for accident analyses or other site factors, such as population distribution are still aids in the overall decision-making process.

It has always been necessary to review the facility / site combination and reach an overall judgment on the adequacy of the balancing of site isolation and facility design.

The current posture was developad some time ago, a representa-tive statement of this position is stated below:21 "In the past, the Ccmission has not licensed large nuclear power plants to be located in or very close to areas of high population density.

In testimony before the Congressional Joint Cormittee on Atomic Energy in 1965, and again in 1967, Commissioner Ramey stated that urban siting required further important advances in reactor plant design, particularly in the capability of safety systems and engineered safety fea-tures, and in adapting these systems so that they can be inspected and tested.

Until these designs and further research and development results are available, and more experience is gained in the design, construction, and opera-tion of large nuclear power plants, the AEC plans to maintain a conservative approach in evaluating plant safety and in establishing a balance between compensating engineering safety features and population density."

4 20For example, Regulatory Guide 4.7 states that special consideration should be given to alternative sites if the cumulative population density exceeds 500 people per square mile at the time of initial plant operation.

The staff has also used, as another rule of thumb, that sites should not be accepted where the surrounding population is greater than the envelope of Zion and Indian Point population distribution.

2I letter, H. L. Price to Senator George Murphy, December 17, 1969.

17 Enclosure "A"

Thus, the established exclusion area, low population zone and population center distances are not the only factors to be considered in the deter-I mination as to whether a selected site is suitable for locating a pro-posed nuclear facility.

The application of the reactor site criteria depends, in the final analysis, on an overall judgment based on consid-eration of 611 factors identified in Part 100.

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ENCLOSURE B DISCUSSION OF CURRENT REGULATORY REQUIREMENTS FOR ACCIDEf!T EVALUATION IN SITING AND LICENSING OF NUCLEAR POWER PLANTS s

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ENCLOSURE B DISCUSSION OF CURRENT REGULATORY REQUIREMENTS FOR ACCIDENT EVALUATION IN

$1 TING AND

'ING OF NUCLEAR POWER PLANTS Introduction There are two principal ar of the NRC staff's review for accident evaluation in siting and licensing of nuclear power plants.

The first aspect of the NRC staff's review is to determine compliance with the siting requirements of 10 CFR Part 100.

Reactor site criteria are estab-lished by Part 100 which in turn, in conjunction with accident evaluations performed by the applicant and the NRC staff, establish boundaries for the exclusion area and the low population zone (LPZ).

In this connection, the NRC itaff has, from the earliest days of licensing recctors, required the use of conservative assumptions and calculational methods in the assessment of consequences from a hypothetical release of radioactive materiaf from a nuclear power plant.

The second aspect of the NRC staff's review is to determine compliance with the licensing requirements of 10 CFR Part 50.

A principal aspect of the NRC staff review is to determine whether the engineered safety features, included in the design of the nuclear power plant to prevent or mitigate a postulated safety system failure or to mitigate the consequences of postulated reactor accidents, can and would perform in a satisfactory manner under normal and abnormal operating conditions.

In this connection, the NRC staff has, from the earliest days 1

Enclosure "B"

't

of licensing reactors, allowed appropriate but conservative credit for the engineered safety features to perform as designed and to mitigate or prevent the system failure and mitigate the consequences of the postulated I

accidents.

1 Current Siting Policy Considerations in Accident Evaluation Practices I

The applicable portions of the Con: missions's regulations which are concerned with siting of nuclear power plants are contained in 10 CFR Part 100.

Section 100.3(a) of Part 100 defines the exclusion area in terms of that

"... area surrounding the reactor, in which the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property from the area."

Section 100.3(b) of Part 100 defines the low population zone in terms of that

"... area immediately surrounding the exclusion area which contains residents, the total number and density of which are such that there is a reasonable probability that appropriate protective measures could be taken in their behalf in the event of a serious accident."

Section 100.3(c) of Part 100 defines the population center distance in terms of f

"...the distance from the reactor to the nearest boundary of a densely populated center containing more than about 25,000 residents."

Section 100.10 of Part 100 states:

" Factors considered in the evaluation of sites include those relating both to the proposed reactor design and the character-istics peculiar to the site.

It is expected that reactors will i

reflect-through their design, construction and operation an 2

Enclosure "B"

\\

extremely low probability for accidents that could result in releases of significant quantities of radioactive fission prod-ucts.

In addition, the site location and the engineered features included as safeguards against the hazardous consequences of an accident, should one occur, should insure a low risk of public exposure."

Section 100.10(a) defines the reactor design characteristics including:

"(2) the extent to which generally accepted engineering standards are applied to the design of the reactor; (3) The extent to which the reactor incorporates unique or unusual features having a significant bearing on the probability or consequences of accidental release of radioactive materials; (4) The safety features that are to be engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur."

Secticn 100.10(c) further defines physical characteristics including:

"(3) Geological and hydrological characteristics of the pro-posed site may have a bearing on the consequences of an escape of radioactive material from the facility.

Special precautions should be planned if a reactor is to be located at a site where a significant quantity of radioactive effluent might accidentally flow into nearby streams or rivers or might find ready access to underground water tables."

The NRC staff, in conjunction with determining the actual distances for the exclusion area, low population zone (LPZ), and population center, performs an independent dose analysis using the actual plant design and selected site characteristics to evaluate the radiatica doses to individ-uals from postulated reactor accidents.

Section 100.11(a) states:

"(a) As an aid in evaluating a proposed site, an appli-cant should assume a fission product releasel from the core, the expected demonstrable leak rate from the containment and the meteorological conditions pertinent j

to his site to derive an exclusion area, a low population L

zone and population center distance."

l l

3 Enclosure "B"

\\

SUPPLEMENTARY INFORMATION:

The U.S. Nuclear Regulatory Commission (NRC) is considering amending its regulations to implement, with respect to NRC and Agreement State licensees, the Agreement between the United States and the International Atomic Energy Agency (IAEA or Agency) for the Application of Safeguards in the United States of America (hereafter called the Agreement).

The amer:dments include a new proposed 10 CFR.Part 75, " Safeguards on Nuclear Material --

Implementation of US/IAEA Agreement," and conforming amendments to 10 CFR Part 40, " Licensing of Source Material," 10 CFR Part 50,

" Licensing of Production and Utilization Facilities," Part 70, "Special Nuclear Material," and 10 ' FR Part 150, " Exemptions and Continued Regulatory Authority in Agreement States Under Section 274."

The amendments will not be issued as final regulations until the Agreement has been consented to by the Senate.

Thereupon, installations determined by the United States to be eligible for application of safeguards under the Agreement, and identified as such on a list to be provided by the United States to the IAEA, will be required to comply with the proposed regulations.

The requirements would not apply to ore processing or to activities at installations which have been determined to have direct national security significance.

The proposed Agreement contemplates that the Agency will be entitled to review certain information for each facility (such as a power reactor or a fuel fabrication facility).on the U.S. eligible 3

Enclosure "B"

\\

"I The fission product release assumed for these calculations should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible.

Such accidents have generally been assumed to result in substan-tial melt-down of the core with subsequent release of appreciable quantities of fission products."

The guideline dose levels used in the applicant's d'etermination are stated as criteria for site areas in Section 100.11(a) of Part 100 as:

"(1) an exclusion area of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

(2) a low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

(3) a population center distance of at least one and one-third times the distance from the reactor to the outer boundary of the low population zone.

In applying this guide, the boundary of the population center shall be determined upon consideration of population distribution.

Political boundaries are not controlling in the applica-tion of this guide.

Where very large cities are involved, a greater distance may be necessary because of total integrated population dose consideration."

If the calculated radiation dose levels are greater than the guideline dose levels stated in Section 100.11 of Part 100, the proposed facility /

site combination does not meet the reactor site criteria.

The site need not be deemed unsuitable for the location of the proposed nuclear power 4

Enclosure "B" 1

I

plant if Section 100.10(d) of Part 100 are met and appropriate engineered safety features can be added to the design to reduce the calculated radia-tion dose levels to less than the guideline dose levels.

The requirements of Section 100.10(d) of Part 100 state:

"Where unfavorable physical characteristics of the site exist, the proposed site may nevertheless be found to be acceptable if the design of the facility includes appropriate and adequate compensating engineering safeguards."

(Should be read as safety features rather than safeguards.)

The site characteristics must meet the requirements stated in Section 100.3 of Part 100 and the distance to the defined areas surrounding the rite must, at a minimum, meet the values derived from the evaluation performed to meet the criteria stated in Section 100.11 of Part 100.

During the site selection process, the applicant may perform many evaluations as stated in Section 100.11 of Part 100 before the actual site is selected on which the proposed nuclear power plant could be built.

These iterations could involve site changes and/or facility design changes before the pre-ferred facility / site combination is submitted to NRC for licensing review.

Further site criteria for multiple reactor facilities are stated in Sec-tion 100.11(b) of Part 100.

Current Licensing Policy Considerations in Accident Evaluation Practices The applicable portions of the Commission's regulations which are concerned with accident evaluation practices in the licensing of nuclear power plants are contained in 10 CFR Part 50.

The interim. statement of policy 5

Enclosure "B"

1 l

i published in the FEDERAL REGISTER on August 21, 1974, summarized the complementary relationship which exists between the regulations in 10 CFR Part 50, for licensing purposes, and 10 CFR Part 100, for siting purposes, regarding accident evaluations.

The policy statement contained the following discussion:

" PROTECTION AGAINST ACCIDENTS IN NUCLEAR POWER REACTORS Interim General Statement of Policy A paramount objective of the Atomic Energy Commission in regulating the design, construction and operation of nuclear power plants is the protection of the public health and safety.

Although the operation of nuclear power plants is not completely risk-free, the safety objective sf the AEC is and always has been to assure that the risk from normal operation and postulated accidents is maintained at an accept-ably low level and to assure that the likelihcod of more severe accidents is extremely small.

The Commission's safety regulations set forth a comprehensive three-level approach to meet this objective.

First, nuclear power plants are required to be designed and constructed with a high degree of reliability so that failures or malfunctions that could lead to accidents are very highly improbable.

An essential part of this first level of safety is the requirement for a comprehensive quality assurance program for plant design, construction, and operation.

The second level of safety is the required provision for measures to forestall or cope with inci-dents and malfunctions that could occur notwithstanding the ascurance offered by careful plant design, construction, and operation.

For example, plants are required to be equipped with reactor protection systems to terminate the nuclear chain reaction quickly and reliably if plant conditions should require such action, and provision is made for leak detection systems to provide indication of incipient fuel cladding failures or degrada-tion of the reactor coolant system pressure bcundary well before leaks become safety problems.

6 Enclosure "B"

The third level of safety is unique to nuclear power plants.

A series of highly unlikely major failures of plant components is postulated as a set of design basis accidents, and safety systems are required to be installed to control all such postulated events.

An example of such a postulated failure is the loss-of-coolant accident used as a design basis for light water power reactors; emergency core cooling systems, whose requirements were recently strengthened in revised regulations (39 FR 1001, January 4,1974),

and containment are provided to mitigate the consequences of such accidents.

The Commission's regulations in 10 CFR Parts 50, "Licens-ing of Production and Utilization Facilities," and 100, " Reactor Site Criter ia," are complementary elements of this third level of safety.

Part 100 requires in effect, that stationary nuclear power reactors be so designed that no design basis accident will result in calculated offsite doses exceeding specified guideline values.

These guideline values are well below levels at which serious injury or death would be expected to occur.

In the approach to safety reflected in the Commission's regula-tions, postulated accidents, for purposes of analysis, are divided into two categories - " credible" and " incredible".

4 The former (" credible") are considered to be within the cate-gory of design basis accidents.

Protective measures are required and provided for all those postulated accidents falling within that category, and proposed sites are evalu-ated by taking into account the conservatively calculated consequences of a spectrum of severe postulated accidents.

Those accidents falling within the " incredible" category are considered.to be so improbable that no such protective measures are required.

x The application of Parts 50 and 100 helps assure public safety by basing protection to the public, both in design and siting, on a very conservative standard for determining and calculating the consequences of postulated potentially severe " credible" accidents.

Part 100 was promulgated at a time when neither the probabilities nor the consequences of these accidents could be calculated with the desired degree of precision.

It was, therefore, considered prudent, as a compensating measure, to require that the consequences of these potentially serious accidents be calculated very conservatively.

Similarly, the acceptable dose guideline values in Part 100 used for evaluating the suitability of reactor sites were conservatively low to com-pensate for any uncertainties in accident analysis.

The draft study, on the other hand, offers a methodology for calculating, 7

Enclosure "B"

as realistically as is now possible, both the probabilities and the consequences of a wide spectrua of accidents, including those considered " Incredible" for purposes of site evaluation under Part 100.

Stated in another manner, under the approach taken in the draft study the distinction recognized in Part 100 between " credible" and incredible" events is eliminated; instead, probability distribution functions for various accident conse-quences are provided."

n n

n Sections of Part 50 which state general licensing requirements on engineered safety features designed to either prevent and mitigate postulated safety system failures or mitigate the consequences of facility accidents include the following:

1.

Section 50.34(a), " Technical information"; and (b), " Contents of applications";

2.

Section 50.36 (c), " Technical specifications";

3.

Section 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors";

4.

Section 50.54, " Conditions of licenses";

5.

Section 50.55a, " Codes and Standards"; and 6.

Section 50.59, " Changes, tests and experiments."

Appendices which state specific licensing requirements for these safety features are the following:

1.

Appendix A - General Design Criteria For Nuclear Power Reactors; 2.'

Appendix B - Quality Assurance Criteria For Nuclear Power Plants and Fuel Reprocessing Plants; 3.

Appendix G - Fracture Toughness Requirements; 8

Enclosure "B"

4.

Appendix H - Reactor Vessel Material Surveillance Program Requirements; 5.

Appendix J - Primary Reactor Containment Leakage Testing F'or Water-Cooled Power Reactors; and 6.

Appendix K - ECCS Evaluation Models.

In the technical specifications appended to the license as required by Section 50.36(c), limiting conditions for operation and surveillance requirements are specified for the performance of specific safety featuree, such as the station containment systems, cleanup systems for the reactor coolant system (specific activity control) and the ventilation systems (air filtration), designed to mitigate the consequences of facility accidents.

Specifications on other safety features designed to prevent and mitigate safety system failures are given as safety limits and limiting safety system settings as well as conditions for operation.

Such safety features include the reactor coolant system, the reactivity control systems and the emergency core cooling systems.

9 Enclosure "B"

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ENCLOSURE C COMPARISON OF ELEMENTS IN ACCIDENT EVALUATION PRACTICE FOR LIGHT-WATER-COOLED REACTORS 4

.W en Hr'

ENCLOSURE C COMPARISON OF ELEMENTS IN ACCIDENT EVALUATION PRACTICE FOR LIGHT-WATER-COOLED REACTORS The attached tables compare the facility design' features (Table I) and the consequence analysis parameters (Table II) for the facility / site combinations approved for licensing over the past 20 years.

The plants labelled "Part 100 Plants" are those facility / site combinations licensed at the time of 10 CFR Part 100 issuance in 196C and which were evaluated in the technical information document, TID-14844, using the published reactor site criteria.

As stated at the time, the guides and the technical document reflected past practice and current policy of the Commission.

The plants labelled " Current Plants" are those facility / site combinations being licensed in the mid-1970's.

The current accident evaluation practices are contained in Regulatory Guides and the Standard Review Plan and are briefly described in Enclosure "A".

For the particular elements of major importance in the accident assessment for LWR's, a comparison is made between typical facility / site characteristics evaluated for establishing the reactor site criteria stated in 10 CFR Part 100 and the typical facility /

site characteristics currently being evaluated against those same criteria.

All of the changes that have occurred in facility design are not reflected in the attached tables.

The changes in design noted have played an impor-l tant role in the changes which have occurred in accident evaluation prac-tices and in characteristics for selected sites over the years.

In the 1

Enclosure "C"

attached tables, the differences between the earlier and current facility design features have been evaluated on the basis of their effect on plant conditions from the postulated accidents, the impact of the changing plant conditions on the plant design and the change in consequences to the public health and safety from all of the interrelated plant accident conditions.

The resulting differences in the parameters for consequence analysir or their values have been discussed on the same basis to provide a common understanding of the interface between plant design features and site related characteristics for evaluating accident consequences.

Several majcr elements important in the accident consequence evalua-tions are depicted in the enclosed tables.

The differences between earliar and current facility / site combinations explain some of the reasons that may have caused the changes indicated in the table.

The following dis-cussion is not intended as a basis for judging the overall value or impact of the changes to the nuclear power field but rather to show the effect on the general risk factor as discussed in the development of the reactor site criteria.

1.

The stated design changes in the reactor core of fuel cladding, fuel performance and core control have resulted in more potential energy in the core during loss-of-coolant accidents but less prompt energy during reactivity insertion accidents.

The total amount of potential energy available during a loss-of-coolant accident in which fuel melt occurs has been increased by nearly a factor of 20 due to the possible metal-water reaction of zirconium and the increased decay heat due to the higher power levels.

The higher burnup and power density are not l

2 Enclosure "C" F

i major contributors to this potential energy increase but do contribute to local fuel temperatures and physical / chemical conditions of fuel and clad.

The changes in control system designs have limited the.

availability of reactivity for accidental 'nsertion, which has reduced the potential prompt energy available to melt (vaporize) fuel during reactivity excursions, and have provided through diversity of nontrol rod actuation and shutdown systems the capability to shutdown and keep a react'or shut cown.

Anticipated transients without scram (ATWS) have a reduced probability of occurrence with current control designs.

2.

The containment system design changes have significantly reduced the contribution of possible shine doses due to any accidental releases.

The added shielding has caused a thermos bottle effect and prevents rapid transfer of any heat produced during a loss-of-coolant accident.

This limitation on heat transfer could result in the predicted high temperatures and pressures evaluated unless other facility design changes were made.

The added concrete containment shield has reduced the probability of vessel puncture and gross failure due to missiles.

The double containment design has further reduced leakage possibili-ties from postulated accidents.

At the time of Part 100 publication, the containment design of boiling water reactors (BWR) was changed to include a pressure suppression system with small containment volumes 3

Enclosure "C"

in contrast to previous designs which were comparable to past and*

present pressurized water reactor-(PWR) designs using large contain-a ment volumes.

The current BWR designs use improved designs of pres-sure suppression with small containment volumes and secondary contain-ment systems while most current PWR designs are using double contain-ment concepts.

3.

The design of engineered safety features have a concern with pre-ventive functions such as emergency core cooling systems (ECCS),

i reactor protective instrumentation (RPIS) and pressure suppression systems as well as a concern with mitigative functions such as atmos-pheric cleanup systems and containment spray systems.

Some of these systems perform a dual function in current facility designs.

The

~

ECCS has become important in the accident scenario because of the l

possible raaid fuel temperature rise in case of loss-of-coolant and 1,

the possible significance of metal-water reactions on the core condi-tions and the atmospheric conditions in the containment.

The RPIS are sophisticated systems that monitor core conditions and primary coolant parameters to detect possible system malfunction and initiate l

other system responses to pretsnt further system failures.

The pressure suppression system is part of the containment system for the BWR and consists of the containment spray system for_ the PWR.

For both major types of water reactors, the systems making up the pres-1 sure suppression system have a dual function.

The atmospheric cleanup l

4 Enclosure "C"

system consists of filters to remove particulate matter and charcoal adsorbers ta remove gases, primarily radiciodines, from the contain-ment atmosphere.

The containment spray system would act to reduce the pressure in the containment and the radioactivity in the contain-ment atmosphere in the event of postulated accidents.

4.

The source term used for the consequence analysis of a loss-of-coolant accident has not changed in the containment atmosphere.

The assumed source term is representative of partial core melt conditions as postulated for the older plant designs but may not be representative of any possible accident condition for current plant designs.

The assumed source term does not contain a release fraction for the radioisotopes of the volatile metals (Cs, Rb, Ru, Rh) or earths (Sr, Ba, La, Ce), the transuranics of the fuel or the activation products in the core materials which could be released under molten core conditions.

However, for a typical postulated loss-of-coolant acci-dent for current core design and operation, the maximum release of radioactive material to be expected would be the radioactivity contained in the fuel rod gap.

Such releases would be significantly less than the source term used in the accident evaluation practice.

5.

The containment leak rates assumed for the older plant designs were probably non-conservative considering both the design and the surveil-lance and testing methods.

The containment leak rates assumed for the current plant designs are probably conservative considering the improved design and accurate testing methods with extensive surveil-lance required for current plant operations, especially for the 5

Enclosure "C" i

4

+

postulated loss-of-coolant accidents without core melt conditions.

As indicated by the WASH-1400 safety study, the current containment design could not prevent uncontrolled containment leakage during certain postulated accidents.

6.

Atmospheric dispersion for the evaluation performed on the older plant designs was representative of average nighttime conditions for the United States and was expected to be that condition or worse 15 to 25 percent of the time.

Current plant designs are evaluated using actual site meteorology for which the dispersion is expected to be that or worse about 5 percent of the time for that selected site.

Even though inany changes have been made and are continuing to be made in the dispersion model, the actual dispersion results t;ve not changed significantly.

The average site meteorology for currently selected sites is probably worse than for the older plant sites.

Dispersion characteristics in bodies of water have not been evaluated since the liquid pathway is not analyzed.

Certain accident cases have used very crude mixing parameters for liquid spills into water-ways but no sophisticated models have been developed.

Liquid pathway studies have recently been completed on a generic basis for off-shore nuclear facilties and land-based nuclear facilities.

These studies have not recommended dose evaluation models but have assessed the possible consequences by such pathways.

6 Enclosure "C"

7.

The dose model used to evaluate the consequences from postulated accidents include the source term and dispersion models previously discussed.

Due to the changes in containment features, the contri-bution to the whale body dose from direct shine has become insignifi-cant to that calculated from the cloud passage and immersion.

Since the thyroid dose from inhalation os the radiciodines contained in the passing cloud was determined using the semi-infinite cloud geometry, the same model has been used for determining the whole body dose from the cloud passage.

This cloud geometry is' assumed to be semi-infinite for both the thyroid and whole body dose model even though the maximum whole body dose can be underestimated for an elevated release and overestimated for a ground level release.

Several finite cloud models are available which could correct this discrepancy.

Only if plant design features are changed such that the whole body could become the critical organ would this aspect of the dose model become important.

The values used to convert the quantity of radiciodines l

inhaled to cumulative thyroid dose have not changed from the values given in 1959 by ICRP.

Current data could shift the overall pre-dicted doses for a typical accidental radiciodine release by less than 10 percent.

As stated previously, releases of radioactive particulates are not assumed from the containment for postulated accidents.

For the older plant designs, the assumed conditions resulted in small quantities of particulate matter being available 7

Enclosure "C" 1

for release from the containmer.t.

The assumption of a filter system (roughing filters and absolute filters) with fallout in the contain-ment and plateout throughout the plant was used as the basis for not evaluating the consequences from released radioactive particulate matter.

An analysis of the release percentage necessary to result in the lung or bone becoming the critical organ under the assumed condi-tions indicated that radioactive particulate matter would not become a controlling factor for the postulated accidents.

However, with major changes in the engineered safety features to mitigate the consequences due to radiofodine releases and the possible releases from core melt or fuel element clad failure accidents, the previous basis for not considering radioactive particulate matter as a major contributor to the consequences may not be adequate.

Only the air-borno exposu e pathway is evaluated by the dose model.

Deposition and rainout models have not been used in the accident evaluation practices.

Such models may not be necessary unless core melt acci-

~

dents with containment failure scenarios are to be evaluated.

4 8.

Only the loss-of-coolant accident, which was considered the con-trolling credible accident, was evaluated against the dose guideline values stated in the reactor site criteria for the older plant designs.

The current plant designs are evaluated for many accident types using dose guideline values related to the relative probability of occur-rence during the lifetime of the plant.

Many of the postulated 4

accidents are evaluated to establish limits in the technical specifica-tions for which plant conditions and surveillance requirements are 8

Enclosure "C"

established.

Such technical specification limits per se were not established for the older plant designs.

For those older plants still operating, such technical specification limits have been established.through an evaluation similar to those performed for the current plant designs.

Potential consequences from many of the design basis accidents can be controlled through restrictions on plant operating parameters such as reactor coolant radioactivity levels, radioactive waste storage levels, containment leakage rates, and removal efficiencies for atmospheric cleanup systems.

In some cases, the loss-of-coolant accident does not represent the control-ling credible accident and other credible accidents with a possibly higher probability of occurrence may be evaluated against the stated dose guideline values.

The restrictions placed on the use of the dose guideline values were:

(1) accident must be credible and rep-resent potential hazards not exceeded by any other accident, '2) accident must have an extremely low probability of occurrence, and (3) consequences for such an accident should represent a low risk of public exposure, i.e., low probability that the predicted dose would be delivered.

For current plant designs, these restric-tions may be somewhat contradictory if the typical loss-of-coolant accident assumptions do not result in the highest consequences from all postulated credible accidents or if the loss-of-coolant accident scenario includes not only core melt conditions but more 9

Enclosure "C"

severe fuel tempt.- dures which can result in containment failure conditions.

Also, because increasingly mora " credit" has been given to e'4gineered safety features, the current consequence evaluation model has become progressively less conservative than the TID-14844 model with respect to the calculated doses at specified distances.

Thus, the built-in safety margin for core melt accidents with con-tainment failure, represented by distance in the TID model, has been reduced by the greater reliance on compensating engineered safety features in the current model.

9.

As stated above, the guideline dose criteria were established on the basis of three restrictions to be placed on the accidents evaluated against such dose levels.

There are no stated guideline dose levels for other types of accidents which do not meet these requirements for determining site suitability.

Since the reactor site criteria as stated represented siting practice for all plants licensed to that date, the licensed plants met the requirements of 10 CFR Part 100.

However, as stated earlier, many other accidents are evaluated for current plant design purposes or restrictions on plant operations which do not relate to site suitability criteria as given in the regulations.

Dose guideline criteria have not been established in the regulations to assess these postulated reactor accidents but staff practices, as indicated in the Standard Review Plan, have been estabitsbed which relate the accident probability relative to the controlling credible accident to establish relative dose guideline 10 Enclosure "C"

6 levels.

The changing plant designs, since the cevelopment of the reactor site criteria, have tended to reduce the predicted conse-quences for the controlling credible accident of the loss-of-coolant accident with some core melt by adding mitigating engineered safety features.

Other plant design changes have reduced the probability of core melt from a loss-of-coolant accident provided the preventive engineered safety features have a minimum acceptable performance under such accident conditions.

Through the combination of these plant design changes, the risks from these accident types where the engineered safety features are effective have been significantly reduced.

Through other plant design changes related to reducing the quantity of radioactive material released during normal operation, i.e., the as-low-as-reasonably-achievable concept, the risks from normal operation and highly probable, abnormal occurrences have been signficantly reduced.

Additional study is required to determine whether the many changes in plant design features, characteristics of site / plant combinations and staff evaluation practices over the past 20 years have resulted in any change in the overall radiological risk from siting and operating nuclear power plants.

Such overall radio-logical risk must include all releases of radioactive material whether from normal operations or postulated accidents along with the probabilities of occurrence.

11 Enclosure "C"

COMPARISDN OF ELEMENTS IN ACCIDENT EVALUATION PRACTICES FOR LIGIT-WATER-COOLED REACTORS (Pt ANT DESIGIS - 1957 10 1977)

TABLE I FACILITY DESIGN FEATilRES Accident Subject Part 100 Plants Current Plants Significance Ispact on Plant Consequence Change Remarks Reactor Core Coseonents Fuel Cladding Stainless Steel -

Zircalloy -

Metal Water Need to control Enhances core melt ilo concern with Zr-il 0 reaction with containment possible metal-Zircalloy clad Stainless Steel Reaction -

2 experimental only in several older (Increased under accident failure (directre-water reactions plants only energy releases conditions.

lease path to environs) at time of duetoZr-ll0).

Part 100/fl0 2

14844.

fuel Parameters

-tow burnup

-liigh burnup

-Illqher release

-More fuel failures

-Release quantity and Impacts on need

[

-Moderate power

-liigh power density percentages

-Higher fuel tempera-quality different.

for quick and density

-Power level up to

-fuel melt times tures - More stored

-Change in critical adequate ECC

-Power level up 4200 Hwt reduced energy organ possible.

capability to 600 b l

-liigher core in-(fuel melt can ventories begin within 30 seconds).

lieactivity Con-Control Rods -

-Control Rods -

Reduce prot.abil-Single system failure ATWS less likely and Ministre con-trol Systems Single System Multi-System Ity for acciden-reduced - ensure reactor chugging fol-quence from

-Liquid Poison tal criticality shutdown.

Iowing certain acci-reactivity type -

System or non-shutdown.

eents prevented, accidents.

$2 8

.i

TABLE I (Cont'd)

FACILITY DESIGN FEATURES Accident Subject Part 100 Plants Current Plants Significance Impact on Plant Consequence Clange Remarks Containment System Primary Containment Steel Vesse!

Steel / Concrete Reduce gross

-Hafntains pressure

-Otract' shine reduced.

Design change Vessel containment with time.

-Cloud passage for caused direct failure pos-

-Decrease gross failure whole body dose dose to be sibilities.

from missile (,enetra-evaluated.

Insignificant tion or overpressare, and failure by puncture un-likely.

Secondary Contain-None to some Douole vessel

-Reduce leakage.

Necessary to cope Reduce releases to Reduces proba-spent confinement with positive

-Increase decay.

with facility design environs - lower bility of gross pressure or pump

-Pressure suppre-changes.

consequences.

containment back capability sfon.

failure due to any postulated accident.

L' Engineered Safety.

Falures-Prevective Emer ency Core Minimum system - no Sophisticated Required to con-Needed to control fuel Prevents significant Due to the v

Cooling Systens review or credit multi-systems -

trol core tem-conditions and pres-releases unless short time from given to prevent extensive review perature and sure increase in con-failure permits coolant loss to core melt, and credit to pre-metal-water re-taltunen t.

core meltdown and initial fuel vent clad failure actions.

significant metal-melt. system and Zr-ll 0 reac-water reactions -

must be rell-tionsal$1mized.

would impact on con-able to prevent tainment.

nejor releases if loss of cool ant should occur.

Ert 8

Ei e

TABLE I (Cont'd)

FACILITY DESIGN FEATtlRES Accident Subject Part 100 Plants Current Plants -

Significance Ispact on Plant Consequence Change Remarks Re:ctor Protective Ministas system -

Sophisticated Reduces probabil-Need to control with Prevents significant Reduced proba-1:strumentation review and credit mult!-systeses -

Ity of abnormal current design of com-fuel temperatures in bility and con-minisal extensive review occurrence re-plex systems.

case of accidents or sequences asso-and credit to sulting in acci-abnormal reactor condl-clated with ah-detept system mal-dent.

tions.

normal reactor functions and pre-operations, vent falluie.

Pressure Sup-Hlnimine system -

Sophisticated Reduces pressure Needed to reduce sus-Reduce driving force Maintain safety pression System no review or mul ti-approaches with time, tajned pressure due to for leakage rate to saargin by com-credit given including spray containment design -

environs.

pensating for systems - PkR and DVR and PWR designs design changes DWR approaches, different.

over the years.

3 ncered Safety E I g,Fea(ures T HitTaative Ateiospheric Clean-Minimum system -

Sophisticated Reduce releases Needed to reduce Reduces thyroid doses.

The improved up System no review or credit multi-systems -

to environs but radiolodine releases, systems are given extensive review ECCS designed to included to and credit to control releases control re-seduce releases from fuel.

leases from an to environs.

accident for which [CCS is designed to control.

Containment Spray Hone Sophisticated to Reduces Needed to control Would reduce airborne Would reduce Systes control alrborne contalianent pres-pressure postulated radioactivity present long term pres-activity and pres-sures from LOCA.

from current designs.

In steam release from sure Lulldup sure buildup.

LOCA.

from radlulysis.

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COMPARISON OF ELEH[NTS IN ACCIDENT EWALUATION PRACTICES FOR LIGitT-WATER-EDOLED REACTORS (F1ANTDESIGNS-1957TO19171 TA8tE II CONSEQUENCE ANALYSIS PARAMETERS Accident Subject Part 100 Plants Current Plants Significance Impact on Plant I Consequence Change Remarks Source Ters Noble Gases 1001 - Fuel 1001 in contain-Fuel melt condi-Controlled leak rate None 100 changes in release (Fr) ment volume tions and delay needed to source even with reduce release to facility design environs.

c%es, 1001 - Con-tainment air-borne (Ca)

Radiolodines 501 - Fr 251 in contain-Fuel melt condl-Mitigation and con-None No change in 255 - Ca ment volume tions trolled leak rate source even with needed.

facility design changes.

N Other I 1-Fr Not considered in Unrelated to Requires filter sys-None Even with fact-1 E-Ca containment.

fuel conditions tem or plate out to lity design control release,

changes, the source has been deleted from con-sideration with-out bases.

Release Rates to Design leak rate - Design leak rate Increase prob-Need to reduce con-Reduce consequences leproved contain-Enviroemient no less than or greater as ability that sequences - Increased fras releases of noble ment design 0.11/ day with specified in leak rate under surveillance and gases and radio-required to meet minimum testing license-testing design condi-sophistication in todines.

Other demands and review.

requirements.

tions(LOCA) testing required.

from facility Extensive review would be less design changes, of design and than design testing specifica-

leakage, m

tions.

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e TA8tE II (Cont'd)

CONSESENCE ANALYSIS PARAMETERS Accident Subject Part 100 Plants Current Plants Significance tapact on Plant Consequence Change llamerks Dispersion Airborne Activity Meteorology used Both 1senersion and Direct shine not Containment shleiding The change from assumed Many changes leave was representa-Inhalation path-considered from has made a clutnge in dispersion conditions been and continue g

tive of about 15 ways. $lte-facility. New need for dispersion to site specific have to be made in the to 20 percent of specific data used modeling is site analysis with addl.

probably resulted in

model, the expected with 5 percentile specific for tional safety features ilttle change to the dispersion for level-models have meteorology.

required by changes to calculated concentra-the U.S.

Only been refined with facility design.

tions in air. Disper-Inhalation con-Improved data for

~

sion characteristics sidered for meteorology.

of selected sites have lodines. Direct probably become worse, shine from con-tainment con-sidered control-Itag for other radioisotopes.

cn Waterborne No analysis No analysis per-None Hone expected Would involve a difft-Only special acci-Activity per formed.

formed but may erent exposure patleday dent cases have be initiated as a and critical organs.

been evaluated for generic issue.

waterborne activ-Ity dispersion.

Dose Model Cloud Geometry haml-infinite Semi-infinite for None Probably none Whole body dose may Finite cloud model for Thyrold-Thyroid and Whole control with specific is available but Direct shine Body.

safety features in not used for whole for whole t.ody.

facility design, body.

$2 Ei LD

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9 TABLE II (Cont'd)

CDNSEQUENCE ANALYSIS PARAMETERS Accident Subject Part 100 Plants Current Plants Significance impact on Plant Consequence Change Remarts Dose Conversion ICRP-1959 dets ICAP-1959 Jata Nor.e None None Current data could cham the factors slightly.

Radioisotopes Nobel gases and Nobel gases and Hone None None except la evalua-Particulates not radiolodines, radiolodine, tion of Ale body doses, used for analysis Others used in Others not con-in environs but direct shine.

sidered, no basis given.

Source Geometry Cloud passage Cloud passage None None Only in whole body Deposition and 3

(Pathways) and direct shine assessment model.

rainout not considered.

Accident Types Maximum Credible Design' Basis Many accident Technical specifica-Consequences may be Even tieugh addt-Accident - some Accidents with types analyzcd-tions limits on plant controlled by technical tional accident fuel melt with assumed fource some are mechan-operation are estab-specifications but types are evalu-containment tenn in contain-Istic others are lished and survell-Ilmiting dose rtill ated the basic

'd intact and ment with specific analyzed for lance set based upon Part 100 guidelines.

guideline dose design leakage leakage rate-system design accident assumptions.

criteria have not rate.

lesser accidents or plant opera-been modified to aIso evaluated tion restrictions include these with no contain-to be stated in accident typs of ment system in license restric-greater prob-release path.

Lions.

atllity. T here Engineered safety has been a con-features func-Linuous decrease t ional.'

in the conserva-tisw included in the analysis as related to the T10-14844 evalua-l tion model, T

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models, given safety margin.

have shifted due to the be drawn by com-Line. By defint-but Part 100 does facility design changes, paring probability tion, these plants not provide the accident evaluation for catastrophic met the Part 100

guidance, modeling has not been WASH-740 (given la accidents reactor site revised.

1957) _(n criteria. The and WASH-1400 basis for the (1975). The pos-s stated guidelines sible change has in Part 100 re-occurred in the late the dose lacreased prob-criteria to the ability of a core low probability of melt accident re-the accident occur-sulting in gross rence and the low containment fall-i probability of de-ure and the re-livering the duced probability stated dose even of a core melt re-if the accident sulting from a occurs. LOCA. i $2 S! Ei 4 1 i i I i