ML19261B971
| ML19261B971 | |
| Person / Time | |
|---|---|
| Issue date: | 02/22/1979 |
| From: | Anderson F NRC OFFICE OF STANDARDS DEVELOPMENT |
| To: | Ahearne J, Bradford P, Gilinsky V, Hendrie J, Kennedy R NRC COMMISSION (OCM) |
| References | |
| FOIA-80-431 SECY-78-624, NUDOCS 7903090220 | |
| Download: ML19261B971 (29) | |
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February 22, 1979 Memorandum For:
Chairman Joseph M. Hendrie Commissioner Victor Gilinsky Commissioner Richard T. Kennedy Commissioner Peter A. Bradford Commissioner John F. Ahearne From:
Fredric D. Anderson, Site Designation Standards Branch, Office of Standards Development Thru:
Executive Director for Operations.M V
Subject:
NRC SITING POLICY AND PRACTICE AN THE PIRG PETITION FOR RULE-MAKING ON POPULATION CONSIDERATIONS (SECY-78-624)
References:
1.
SECY-78-624, "Public Interest Research Group, Et al., Peti-tion for Rulemaking to Amend 10 CFR Part 100 Pertaining to Population Density Criteria Around Nuclear Reactor Sites,"
(Specifically Enclosure "K"), December 4, 1978.
2.
Memorandum, " Comments on SECY-78-624--PIRG Petition for Rulemaking on Population Density Around Reactor Sites,"
James L. Kelley, OGC, and Kenneth S. Pedersen, OPE, to Commissioners, January 4, 1979.
3.
Memorandum, "Requect for Progress Briefing by the Siting Policy Task Force,
-muel J. Chilk, Secretary, to Lee V.
Gossicx, EDO, Janc.y 5, 1979.
4.
Memorandum, " Comments on SECY-78-624--PIRG Petition for Rule-making on Population Density Around Reactor Sites," John Ahearne to K. Pedersen, OPE, and J. Kelley, Acting GC, January 9, 1979.
5.
Memorandum, " Comments on SECY-78-624--PIRG 'etition for Rule-making on Population Density Around Reactor Sites," Samuel J.
Chilk, Secretary, to Lee V. Gossick, EDO, January 11, 1979.
6.
Memorandum, "SECY-78-624--PIRG Petition for Rulemaking on Population Density Around Reactor Sites," Howard K. Shapar, ELD, thru Lee V. Gossick, EDO, to Commissioners, January 11, 1979.
79030907&O
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7.
Commission Briefing, " Progress by Siting Policy Task Force and on SECY-78-624--PIRG Petition for Rulemaking to Amend 10 CFR 100 Pertaining to Population Density Criteria Around Nuclear Reactor Sites," January 18, 1979.
Purpose:
To provide the Commission a more detailed report on the subject issues as discussed during a recent staff briefing (Reference 7).
Copies of this report are to be provided to the NRC Siting Policy Task Force.
Background:
I have been involved in the development of reactor site criteria since joining the regulatory staff in 1961. My first assignment was reviewing and revising a draft of the technical document, TID-14844, " Calculation of Distance Factors For Power and Test Reactor Sites" which was being prepared to support 10 CFR oart 100, " Reactor Site Criteria." As adjunct to this assign-ment, I was directly involved in the publication of the regula-tion, which included an understanding of the basis for the require-ments on site suitability given in 10 CFR Part 100. Although frequently contacted by individual regulatory staff members regard-ing the " meaning" or rationale of 10 CFR Part 100 requirements and TID-14844 assumptions, I was not directly involved in the implementation of current siting policy for licensing nuclear power plants after 1962.
In the mid 1960's, I contributed to the development of design and supplementary criteria which later became Appendix A to 10 CFR Part 50, " General Design Criteria," and some of the earlier regu-latory guides.
The initial concept of ALARA criteria as stated in Appendix I to 10 CFR Part 50 was proposed by several of us at that time.
I also contributed to the development of accident evaluation models and assessment of population criteria, includ-ing consideration of population distribution concepts - some regu-latory gaides were developed from these reactor technology memo-randa. These efforts were primarily related to radiological aspects as stated in 10 CFR Parts 20 and 100.
In 1968, I was assigned to operating reactors as a Project Manager for licensed facilities and as a Radiation Specialist for the radiological assessment of operating reactor license amendments.
I contributed to the development of accident evaluation models for specific design basis accidents and to the generic review of several engineering items important to operating reactor designs.
The operating nuclear facilities which were assigned to me included critical facilities, research reactors and test reactors as well as power reactors.
I was Project Manager for the decommissioning of various types of nuclear facilities including several power reactors.
I was directly involved in the implementation of siting
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policy issues in the area of radiological consequences from normal operations (Appendix I to 10 CFR Part 50) and from postulated reactor accidents for operating reactors as well as the day-to-day overview of licensing policy issues for specific operating reactors.
Most of this effort was directed toward the amendment of technical specifications for the specific facility.
In 1975, I was assigned to the Office of Standards Development to become the SD task leader and principal investigator on the project to review reactor siting policy and practice for the Commission and, later, the SD task leader for the PIRG petition for rulemaking (PRM-100-2) on population density criteria.
This review of siting policy and practice has resulted in a total of twelve Commission papers, which have discussed past, present and proposed siting policy, regulations and staff practices for vari-aus issues. These papers were prepared in cooperation with NRR, NMSS, RES, SP and ELD.
In, addition to the general review of sit-ing policy and practice issues, I have directly contributed to the evaluation of the specific issues regarding emergency planning and preparedness and accident evaluation practices.
I have kept informed on the staff developments in the specific areas of alternative site evaluations and procedures, including the geologic and seismic review policy and practices.
In order to coordinate this effort with the international effort in the reactor siting policy and practice area, I have been the 50 task leader for the coordination of NRC review of the parallel IAEA effort on IAEA Code of Practice on Safety In Siting and related IAEA Safety Guides.
I prepared a proposed revision to 10 CFR Part 100, which was patterned after recent standards work with the IAEA, in SECY-77-288, dated June 7, 1977.
Some of this siting philosophy has been incorporated into the proposed rule for licensing storage of spent fuel in an independent spent fuel storage installation, 10 CFR Part 72, and may be incorporated into the proposed rule for geologic repositories, Subpart B to 10 CFR Part 60, " Disposal of High Level Radioactive Wastes In Geologic Repositories."
Discussion:
In response to requests and comments in References 2 through 6 and to supplement my camments in Enclosure "K" to Reference 1, I will provide some f act. on " population concepts" for siting policy and on the historical ;evelopment of current staff practices (cause and effect), as well as my concerns with the four policy issues discussed in 'leference 2 and with the three concepts men-tioned in Reference 3.
I will address the cuestions raised in References 4 and 5 and the implication in Reference 6 (that is, Part 100 is an ineffective and antiquated regulation for siting).
This information report is in response to my offer at the Commis-sion meeting on January 18, 1979, to provide a more detailed report
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of the issues discussed in my briefing (Reference 7), if it would be of value to the Commission in their deliberations on siting policy and practice.
A request was made by Commissioner ".ennedy that I prepare such a report for the Commission and provide the Siting Policy Task Force with a copy.
Siting Policy On Population Concepts The establishment of the three basic siting terms in 10 CFR Part 100 had the purpose of defining specific areas and distances to be used in determining site suitability for a power or test
-All of these terms are related to protecting the health reactor.
and safety oT specific populations within the general public.
1.
E(clusion Area - defined to limit access and ensure licensee control of activities within the area.
The reference dose levels for a two hour exposure at any point on the exclusion area boundary define the size of the area and restrict the postulated releases of radioctive material from the reactor.
Most licensed facility /
site combinations (75 percent) are limited by the doses calculated at the exclusion area boundary. The control established by these definitions for the exclusion area is intended to protect the onsite personnel from excessive exposures during the period of time required to take protective action under postulated accident conditions.
Establishment of permissible postulated exposures of personnel in the control room (General Design Criterion 19) is an extension of this philosophy.
Therefore, a minimum exclu-sion area distance would have to be accompanied by a minimum credit level requirement for engineered safety features to maintain the same or lower risk level to onsite personnel as defined by Part 100 requirements.
The exclusion area concept was not intended as a site suitability feature related to protecting offsite population except as a secondary effect due to limiting releases of radio-active material from the reactor.
2.
Low Pooulation Zone - defined as a protective action area with limitea population levels.
The reference dose levels for the cloud passage exposure time period at any point on the outer boundary (inner boundary is exclusion area with its reference dose level criteria) defines the size of the area and restricts the postulated releases of radioactive material from the reactor.
The control established by tnese definitions for the LPZ is intended to protect the nearby offsite population from excessive exposures during the entire period of time (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> up to entire cicud passage time) during wnich releases of radioactive material to the environs may occur under postulated accident conditions.
Most licensed facility designs would limit the time period asso-ciated with entire cloud passage exposures of significance to
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less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
A minimum low population zone outer boundary distance would have to be accompanied by a minimum credit level requirement for engineered safety features to maintain the same er lower risk level to the population within the LPZ as defined by Part 100 requirements.
The low population zone concept was not intended as a site suitability feature related to population distribution and density aspects except as a secondary effect due to requiring reasonable assurance that protective actions could be taken to further reduce the risk to the health and safety of the population within the LPZ from postulated accidents.
The risk reduction for such populations due to protective actions that could be taken under postulated accident conditions is not a factor in the Part 100 criteria for determining site suitability.
3.
Pooulation Center - defined as densely populated center of more than aoout 25,000 residents which may represent a group too large for effective protective action a be taken.
Rather than reference dose levels, a distance factor related to the distance to the LPZ outer boundary defines the minimum distance required to reduce the postulated integrated population exposure to such a group of people below which protective action is not necessary even from a catastrophic accident resulting in large releases of radioactive material.
For larger cities, a greater distance might be warranted to satisfy the consideration of total integrated population doses.
The population center concept was not intended as a site suitability feature but rather to satisfy a basic objec-tive to ensure that the cumulative doses to large numbers of people should be low as a consequence of any nuclear accident even a highly improbable accident not considered credible.
The total population dose concept was to be studied further to refine this requirement. The population center concept was intended to limit the risk to the health and safety of the public from any nuclear accident regardless of the magnitude of the releases.
Site Reviews and Staff Practices Three issues important to site reviews, for which staff practices have been developed to implement Part 100 criteria, are (1) popu-lation criteria, (2) accident evaluation assumptions, and (3) protective features.
The staff review practices have contributed to the changes in site reviews.
The historical development of these issues with associated consequences are discussed below.
1.
Poculation Criteria - The relation = hip between the Part 100 siting terms ano specific population categories have been dis-cussed.
The purpose of the exclusion area has not changed even though staff understanding of the need for an exclusion area as
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a protective feature for onsite personnel has been lost.
- However, a more restrictive definition on licensee control has been imposed through staff practices by requiring ownership of the exclusion area by the licensee.
The definition for low population zone arm the related population center have been modified by staff practices to include pooulation density considerations.
Population density criteria are used in staff reviews of sites as discussed in SECY-78-137 and SECY-78-624 and stated in Regulatory Guide 4.7 and Standara Review Plans.
Based upon the results of staff analysis for integrated population doses at approved sites, a predictable correlation between popula-tion density as used in staff practice and potential integrated population exposures does not exist.
These results were obtained from the Final Environmental Impact Statements for a typical loss-of-coolant accident (large pipe break condition) using the specific site / plant related characteristics.
Site factors such as population distribution within the area, dispersion characteristics and topog-raphy, as well as the plant's engineered safety features can more than compromise the influence of population density on the predicted impact or consequences from potential accidents on the population or environs. For example, sites with similar population densities have calculated integrated population doses that vary by more than a factor of 100.
For the 94 site / plant combinations examined, no predictable relationship between the calculated integrated population doses from postulated accidents and the site specific population density levels was observed.
The magnitude of these doses is not indicative of the impact on the population or environs but can be used as a comparative tool to indicate the lack of meaning in population density levels for evaluating site suitabil-ity as used in current staff practice.
Thus, current staff prac-tice regarding use of population density criteria is contrary to the purpose of siting terms defined in Part 100 and may be counter productive in the determination of site suitability for a given nuclear power plant design.
Another staff practice, which is contrary to the purpose of the basic siting terms, is the use of the reference dose levels as absolute acceptable dose limits.
The reference dose levels stated in Part.100 were established as guidelines tn assess the suitabil-ity of the site dimension and distances for locating the precosed reactor design, not as absolute limits. Absolute limits would have implied that if the calculated doses were less than the stated levels, the site was acceptable, but if the calculated doses were greater, the site was unacceptable.
The reference dose levels stated in Part 100 were acpropriate for assessing site suitability for a given reactor design if (ana only if) the purpose of the
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criteria and the method described for assessing the site distances were used.
The current staff practice does not imolement the purpose of Part 100 site criteria.
The impoFtint variaole against which the reference dose levels were to be compared was the dis-tance value for the exclusion area and the distance value for the low population zone outer boundary.
The staff judgement to be made in determining site suitability was whether the site distances calculated for the reactor design using these reference dose levels were comparable to those of the proposed site. All of the stated site evaluation factors were to be considered in the staff determination of acceptability of the site / plant design combination.
The site distance factor was only one of many site factors which were to be used in the weighing of site / plant design acceptability--not as a pass-fail criterion.
2.
Accident Eval _uation Assumotions - The accident ev.luxcion to be performed for'Part 100 is defined in the regulation as "a major accident that would result in potential hazards not xceeded by tt,ase from any accident considered credible." This accident was defined as the maximum c edible accident (MCA), which was postu-lated to be a LOCA-large pipe break and was evaluated in TID-14844.
Contrary to the stated purpose of these criteria, staff practices have changed the scope to assess the consequences from other postulated accidents under the criteria stated in Part 100. These postulated accidents were ambiguously designated as design basis accidents (C8A) and replaced the MCA in the staff accident evalu-ations.
The postulated LOCA has remained as one of the DBAs and is evaluated in the same manner as the original MCA.
This LOCA-large pipe break accident is postulated to be equivalent to a core melt accident with containment intact. A proposed rule (annex to Appendix D of Part 50) included postulated accidents ranked by classes for environmental impact evaluations--a proposal which confused the accident evaluation issue beyond explanation.
From this Commission issuance came the term " Class 9 accidents",
which has been confused by the staff and public alike with core melt accidents of the MCA class and, more recently, WASH-1400 core melt accidents. Assumptions to be used with a Class 9 acci-dent were never defined in the preposed rule because the accident was judged to represent a low environmental risk.
However, all assumptions associated with the evaluation of environmental impact are to be " realistic" rather than conservative as used for safety evaluations.
The use of the term. Class 9. with the conservative assumptions of a core melt acc1 cent is inconsistent and confusing to all carties. One or the purposes of Part 100 was to encourage the tecnnological development of engineered safety features so that siting distances could be reduced.
This purpose has been realized by the improved design cnaracteristics of engineered
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safety features and by the changing staff practices in assessing credit to be allowed for the engineered safety features in cur-rent reactor designs.
3.
Protective Features - The early reactor sites were considered suitable for locating the reactor designs of the 1950's on the basis of isolation from people or the distance factor.
- However, as discussed above, the Part 100 reactor criteria encouraged the development of engineered safety features to reduce the reliance on distance for protecting the health and safety of.the public.
For example, the use of the TID-14844 calculational assumptions for engineered safety features with current reactor designs would result in as much as a factor of 10 increase in distances for the exclusion area and LPZ as compared to the approved sites' distances for these siting terms and a factor of 100 increase in the areas that would be enclosed by these boundaries.
If this development continued to its ultimate conclusion, a future reactor desion could be located anywhere, including an area within a metro-politan center, and the combination site / plant design would be considered acceptable and licensable.
Therefore, the staff prac-tices for evaluating the credit to be allowed engineered safety features has resulted in the observed siting concern with what is an acceptable balance between distance as a protective site feature and engineered safety features as a protective plant feature.
4.
Staff Review Practices - Staff reviews in the 1960's were performed by a few staff reviewers with general safety and plant design knowledge.
Such reviews were not detailed and were based upon the experience and judgment of the individual staff reviewers.
I call these reviews " customized" to the plant / site combination.
In the 1970's, the staff review procedures were revised in order to process a large number of applications on similar reactor designs.
The current reviews are performed by many reviewers with only specialty interests, not general--and are very detailed with check list criteria.
The criteria and judgment of the indi-viuual reviewers have been replaced by standard review plans and regulatory staff positions.
This review procedure eliminates consideration of special circumstances by staff reviewers and reduces the involvement of the review staff in any given licens-ing decision.
It provides for a uniform review of individual systems without a means for balancing the weaknesses and strengths of the overall plant design.
I call these reviews " assembly line evaluations of the plant / site combination.
Essentially only the name of the plant / site needed to be changed in some of the staff's safety and environmental reports.
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The question is what type of staff reviews should be performed in the 1980's? Should the reviews become a matter of checklist criteria and computer results? This approach is being promoted by such actions as the codification of current staff practice on population density.
In such reviews, staff experience and judg-ments are completely eliminated and all plant safety design improvements may be frozen.
An independent staff evaluation of the site / plant design combination will not be made since the staff review will become a matter of comparing the contents of the application with the " acceptance" criteria.
The only check and balance remaining would be the inspections performed at the site during construction, startup of the plant and operations.
Referenced Policy Issues and Conceots The polic.y issues and concepts raised in References 2 and 3 by the Offices of Policy Evaluation, General Counsel and the Secre-tary are indicative of the misconception and lack of knowledge in the staff regarding the siting policy reflected in the regula-tions.
The siting policy which was used as a basis for the guide-line criteria published in the regulations was discussed in the Statement of Considerations accompanying the promulgation of 10 CFR Part 100 and in the technical information document.
TID-14844.
Staff practices, standard review plans and regulatory guides have contradicted the purpose and scope of the regulation and the underlying policy. One of the major issues for Commissicn resolution is:
Should the siting policy as reflected in our regulations be amended to conform with staff practices (published or Should the as regulatory guides and standard review plans) lur regulations staff be reminded of the siting policy stated in followed by modification of staff practices? A third option to resolution of this issue would be:
Should a complete review of siting philosophies, regulations and staff practices be performed to derive & consistent siting policy between regulations and staff practices.
This last option may be the ideal solution, but it is neither realistic nor attainable.
However, I understand from the comments contained in References 2 and 3 that this option would be the one recommended by your staff Offices.
In the following discussion, I have commented on the particular four issues stated in Reference 2 and the three concepts raised in Reference 3.
For convenience, I have restated the issues as given in these references prior to giving my comments.
1.
Codification - Should policy be codified in regulations?
Comment:
The regulations should reflect the siting policy and do incorporate the policy but not staff practices as I have outlined
The Commissioners 10 in the previous discussions.
I believe that only technically sound staff practices should be codified and such practices must be consistent with the siting policy reflected in the regulations.
Current staff practice regarding population density considerations does not meet either of these criteria. Although the present siting regulation (10 CFR Part 100) is not written as a regulation but as a siting guide, the policy it respresents is not vague and is not out of date.
The fault, if any, rests with the ambiguity between the policy as written in the regulation and the way staff practices have further served to confuse the siting issues.
2.
Conceot of Exclusionary Limits - Should there be firm, exclu-sionary limits, flex 1ble guidelines or some combination of both?
Comment:
If one examines the meaning of the siting terms as. I have discussed them, this concept becomes meaningless unless the policy is changed or the level of associated risks to individ-uals and to the population is to become more variable than is currently the case. A further examination of what a Class 9 acci-dent means and how it can be related to accident evaluation policy in the regulation is required. My position would be to remove the proposed annex to Appendix 0 of 10 CFR Part 50 from the Federal Register Notices and from Regulatory Guide 4.2 and, instead, estab-lish a staff practice for evaluating accidents in a realistic manner for both safety and environmental impact.
I have drafted such a regulatory guide to replace Regulatory Guide 1.4 and another Part 100 related accident evaluation regulatory guide to replace Regulatory Guide 1.3.
These drafts were developed as part of the followup paper to the staff paper on accident evaluation prac-tices (SECY-78-111).
This proposed paper is to be included in the overall siting policy review being performed by the NRC Siting Policy Task Force.
3.
Parameters on Which Limits should be Placed - Should the required degree of reactor site isolation in Part 100 be specified in terms that include?
a) limits on population density out to a specified distance; b) minimum exclusion distances; and (c) minimum distances to " low population zone" boundaries?
Comment:
There is no merit in minimum exclusion or LPZ distances or in maximum allowaole population density levels as I have dis-cussed previously. These suggestions are contrary to the siting policy reflected in the regulations.
This concept for siting
The Commissioners 11 philosophy would require not only new regulations but also new siting policy and staff practices. This approach would lead to the third option requiring a complete review of siting philosophies and a major upheaval of past siting decisions without an apparent purpose or benefit.
4.
Degrea of Conservativeness - Should there be a policy shift in the ccnservative direction?
(I.e., should NRC in the future require a greater degree of isolation than that of the least isolated of the already approved sites?)
Comment:
Due to the misconception of your staff Offices regarding the siting policy incorporated into the regulations (criteria of Part 100), the use of limits on distance and population density parameters are viewed as increased conservatism.
In reality, as I have previously discussed, the result would be an increased risk to the affected populations unless other restrictions were included in the proposed regulations to maintain the level of safety afforded by the Part 100 guidelines.
The importanca of balancing all factors affecting the safety of a given plant design /
site combination must be understood to prevent arbitrary recom-mendations (which appear to improve safety), which result in reducing overall public health and safety.
5.
Firm, exclusionary limits-for example, with respect to upper bounds on nearby population levels or on seismic characteristics.
Comment: This concept in a general siting policy would require extensive investigation to ensure that such limits were technically sound and would maintain or improve the level of public health and safety obtained by the use of the Part 100 criteria and guide-lines. The previous discussion of this concept presented above would imply that the concept would not be an improvement aid could be a detriment in the siting policy '-
6.
Limits that, though not exclusionary, would trigger a presump-tion of " obvious inferiority" in site comparisons.
Comment: This concept may be more destructive in a general siting policy than the proposed exclusionary limits due to the favored position such site parameters would have in a site suitability evalcation. Without a balancing of all site parameters which affect safety or envircnmental impact, parameters unimportant to the suitability of a given site for a given plant design could becoma controlling.
Also as a trigger, the parameter " limit" woulo not have the peer re/iew required for an exclusionary limit in a regulation to assure that a check and balance existed to determine its importance.
t The Commissioners 12 7.
The threshold concept; i.e., whether some relatively insignifi-cant impacts may be properly neglected in site evaluations.
Comment:
This concept is inherent in most staff review practices and is inferred by the guidance stated in Part 100 criteria.
The threshold concept has been a part of siting policy, regula-tions and staff practices for many years. Whether the insignifi-cant impacts can be identified and an absolute threshold level established for general application (without considering inter-related factors that influence the importance of the impact from plant / site to plant / site combination) is technically remote.
Ouestions on Petition Response My response to the questions raised in References 4 and 5 was submitted as part of my differing staff view in Appendix "K" to SECY-78-624 and my previous discussion of population criteria under the enclosed section, " Site Reviews and Staff Practices."
I do not believe that specific numerical criteria can be supported for population density levels.
Such criteria would not be consis-tent with the siting policy represented by the current regulations.
The weighting of transient populations by time in an area is appropriate in a risk type assessment; it would not be appropriate for evaluating either the consequences to an individual or the adequacy of an emergency plan to provide reasonable assurance s
that protectiu actions could be taken.
Therefore, there is no single argument for tha accroariate treatment of transient cooula-tions.
The use of population censity levels or population center distances in a site suitability analysis would be dependent upon the purpose to be accomplished by the use of the given parameter.
Current regulations which reflect a given siting policy would argue for the use of population center distance as the proper parameter but a different siting policy could argue for the use of population density levels.
As I have stated before, if a siting policy different from that used to establish current regu-lations (guides) is promulgated, the arguments will change.
Part 100 As A Siting Regulation As I have stated several times in this discussion, Part 100 is written as a guide (and an interim guide at that)--not a regulation--
which states the site criteria used by the Commission (AEC) to evaluate site suitability for power and test reactors.
It does represent a given siting policy based upon past experience of siting known reactor types.
The previous discussion does state the purpose of the basic siting terms in Part 100 while the State-ment of Considerations provided some of the philosophy associated with the guides for evaluating site suitability factors.
Another
The Commissioners 13 basic fact of the policy which prompted the development of Part 100 guidance is that the suitability of a site was to be directly related to the plant design proposed for the site location.
There could be no separation of the decision of site suitability from the knowledge of the licensability of the plant design / site combi-nation.
Recent staff evaluations have advocated the need to separate site suitability determination from the plant design or licensability of a standard power plant.
This concept is a circumvention of basic siting policy developed for the promulgation of Part 100 criteria and 9 art 50 licensing requirements.
These questions have occurred as a result of the promotion of standardized power plant designs (components and systems) with early site review options. Development of new concepts such as these by the staff never considered the impact on established siting and licensing policies.
I believe this resulted from a lack of knowledge by the staff promoting such concepts of the policy that established the siting and licensing regulations.
I do not believe that Part 100 is an ineffective regulation since we have licensed some 100 plant / sites under those provisions or that the siting policy reflected by Part 100 criteria is antiquated.
I do believe that Part 100 should be rE-vised to reflect siting requirements as a
.egulation rather than as guides is currently written. As stated earlier, I have drafted a revised Part 100, which was Enclosure "E" to SECY-77-288, dated June 7, 1977, that would, in fact, main-tain the past siting policy in regulatory form.
The problem with siting policy is not the criteria and guides as expressed in the regulations (Parts 50 and 100) and in Statements of Considerations that accompanied their promulgation but in the developing and changing staff practices which have been used to implement a siting policy that was not understood by the licensing staff.
I continue to stress the need by the staff to understand the scope, purpose and basis of any staff practice before the staff uses such practices under different circumstances.
Time has eroded the siting policy used to develop the regulations while the staff practices used by the licensing staff in its reviews and approval of reactor plant design / site combinations have been changed without an understanding of the basic regulations.
This situation has been complicated by the assembly line type of safety and environmental reviews performed by the licensing staff.
Whereas Part 50 has been a dynamic (although cumbersome and volu-minous) regulation for licensing facilities as indicated by its many amendments since 1962 (when Part 100 was promulgated), Part 100 has been a static regulation for siting these facilities.
Part 100 has added one appendix on seismic and geologic criteria,
The Commissioners 14 which is under serious challenge because of its inadequacies, and a minor change to the determination of a population center, which is inconsistent with the initial intent of the criteria as discussed previously.
Relationship of Siting Policy and Regulatory Organization Development The following discussion provides my observation of the relation-ship between siting policy knowledge in the staff and the develop-ment of the regulatory organization from promulgation of Part 100 (1962) to the present. As discussed under the subsection, " Staff Review Practices," the staff review has changed from a " customized" licensing review in the 1960's to an " assembly line" licensing review in the 1970's which was due to the increased environmental review for NEPA and the increased number of license applications.
Most of the technical review staff (defined as project managers and branch chiefs) during these earlier project reviews have either left the regulatory organiza; ion or been promoted to the managerial level (defined as assistant directors and above).
These staff members were familiar with the staff practices and procedures used to implement the siting policy promoted by the Part 100 criteria. The current technical review staff is familiar with the current staff practices and procedures but are not knowl-edgeable in the siting policy that established the regulations.
Therefore, with few exceptions, the Commission has a managerial staff that is familiar with the staff practices used at the time of the promulgation of the regulations but not current staff practices, while the technical review staff is familiar with current staff practices but not the regulations or the development of staff practices.
I do not believe that a mutual understanding of the problems with siting policy and regulations on the technical review staff side and with current staff practices on the managerial staff side exists within the regulatory staff.
I do not have a solution for resolving this communication gap other than to estab-lish a task force of cognizant staff members, bo a managerial and technical review staff members, to review anc revise the com-bined facility licensing and siting regulations and establish consistent staff review practices.
I do believe that much of the technical review staff unrest has been caused by the misunder-2 standing that exists between current staff practices and the siting and licensing policies as expressed in our regulations.
==
Conclusions:==
1.
All regulations which relate to the facility licensing pro-cess should be reviewed as a package and revised to reflect a cohesive and consistent oolicy.
These regulations include, at a minimum, Parts 20, 50, 51, and 100.
One cannot separate
The Commissioners 15 siting policy as reflected in Part 100 from the companion facility licensing policy.
2.
Task Force should be established to perform such a review.
Its membership must include knowledgeable individuals in the various disciplines essential to reviewing the regulations--regar.dless of their position / level in the current NRC organization.
Peer review becomes the most important consideration in such an endeavor.
Limit the number cf members (less than 10) and give the Task Force the authority and responsibility to rewrite the regulations and to estab-lish purpose, scope and basis for a uniform facility licensing process.
The current Siting Policy Task Force does not meet any of the above requirements and the associated Working Group members meet few of the needed qualifications.
3.
The current staff practices as reflected in the Regulatory Guides and Standard Review Plans do not implement the siting policy or criteria stated in the regulations.
In many cases, the staff practices are not only in conflict with the regula-tions and siting policy but are in conflict with each other.
These internal conflicts are most prominent in a comparison of environmental and safety evaluation practices.
4.
The current regulations are not antiquated or ineffective.
Because the current staff practices as reflected in Regula-tory Guides, Standard Review Plans, Safety Evaluation Reports and Environmental Impact Statements are inconsistent with current regulations does not necessarily mean that the regulations are in need of revision. Unless a basis can be established to justify current staff practices, I would support (1) the retention of the regulations which do have a basis and (2) a revision of staff practices to implement the regulations and their siting policy.
Recommendations:
1.
Current staff practices on population concepts should be changed to implement regulations as written and intended by siting policy.
2.
The facility licensing staff should be directed to develop meaningful criteria for a common review of population con-siderations for safety and environmental impact evaluations.
The Commissioners 16 3.
The staff, i.e., NRC Siting Policy Task Force, should be directed to develop the purpose for each siting policy criteria identified and the means for implementation in the regulations or staff practices.
4.
The EDO should be directed to establish a task force of cog-nizant staff members regardless of organization or position to review and revise the combined facility licensing and siting regulations. The makeup of the AEC Regulatory Staff Task Force used to evaluate the Draft Reactor Safety Study Report, WASH-1400, could serve as an example.
Su cf
,?
eu Fredric D. Anderson Site Designation Standards Branch Office of Standards Devalopment cc:
SECY PE GC CA IA DA EDO AD" ELD
""SS IE RES SP MPA Siting Policy Task Force " embers
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D+ s Lc-r r January 4, ic79
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J EEMORANDUM FOR:
Chairman Hendrie r
Commissioner Gilinsky Commissicner Kennedy
/
Commissioner Bradford Commissioner Ahearne
(
l FROM:
Kenneth S.
Federsen, Directcr
.,0ffice of Policy E.aluation '
d James L. Kelley Acting General Counsel 4
S URJECT:
COMMENTS CN SECY-78-624 -- PIRG PETITION FOR RULEMAKING ON POPULATION DENSITY AROUND REACTOR SITES A.
OVERVIEW Since 1975 the Ccmmission has had an ongoing effort aimed at pro.ducing an explicit and cohesive updated siting policy.
The effort has recently received new impetus through forma-tion, at Commission request, of the Siting Policy Task Fcree, an NER-led multi-office staff study group, to review and study siting policy ard formulate -- for Ccmmission consideration -- a cohesive, specific, and definitive statement.
The PI?.G petition en pcpulation density criteria addresses scme of the mes: tasic issues involved in the engoing policy developmen efforts.
Current reviews (by the Task Force and c:her staff groups) include general siting policy, emergency planning pclicy, and policy en alternative site evaluations, all of which are relevant to the PIRG petition.
The petitioners assert that their pcsition " accords with current NRC official golicy against high pcpulation density Centacts:
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Reference 2
The Ccnnissicn January h, ic79 siting."
(Encl.
A, p.
3)
They argue tha: tha: pclicy has nct been applied riscrcusly encush and that the AEC and NRC "have permitted frecuent breaches" of it.
(Encl.
A, p.
3)
They ask that what they perceive as a iccse, flexible policy, not formally documented in specific terms, be put in regu-latory form and that it be modified to forbid siting where population exceeds specified numerical limits.
(Encl. A, pp. 3 and 6)
The petition includes a specific set of pro-posed amendments to the siting regulations.(10 CFR 100).
(Encl.
A, pp. 21-23)
The staff's propcsed disposition of the petition (Encl. H) would grant in part PIRG's recuest to put policy into regu-latory form, but would ceny r1RG's proposals on the basis of the pr t=ed adequacy of present Part 100 and staff practice.
This ignores the fact that it is generally agreed that Part 100 is in need of extensive revision.
In fact, the Siting Policy Task Force is currently probing overall siting policy with such pessible revision in mind.
The implicit defense of Part 100 should not be used while that same policy is undergoing study for the purpose of potential comprehensive change.
Adopting clear numerical criteria for population density incorporating some of the basic features cf the PIRG proposal certainly appears to be an ap.rc.oriate encuch policy option to warrant evaluation in the ongoing policy study -- withcut the adverse presumption suggested by the paper.
Our cc =ents en an approach to the paper center on two themes developed in the folicwing two secticns.
Before the Cctmission can give prcper attention to the PIRG p etition, it needs to address four maj or policy considerations.
Apart frcm policy develc,cnent censiderations, the proposed response to the peticien is deficient in scoe important specific respects.
Accordingly, C?E and CGC recccmend that the Cc missicn (a) no:
approve the prcpesed respcnse as new written, (b) seek advice frem the Sizing Policy Task Fcree, and (c) receive a staff briefing bercre passing on the nerits of the ?1?.0 petition.
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two-and-a-half years.
.S.dmittedly, cur suggestions will delay acticn further.
We have reluctantly concluded, however, that scme delay is warranted to produce a straightforward, cohesive responsive document that gives adequate attention to the deficiencies extant in N?.C regulations.
3.
POLICY ISSUES Disposition of the ?IEG petition by the Cctmission requires addressing four basic policy issues with respect to popula-tion levels nr 1r nuclear pcwer plants:
1.
Shoul-policy be codified in egulations?
2.
Should there be firm, exclusionary limits, flexible guidelines, or sc=e ccabination of both?
3 Should the required degree of reactcr 4.t e isolation in ?ar: 100 be specified in arms that include:
(a) limits en population density out to a specified distance; (b) minimum exclusion distances; and (c) minimum distances to " low pcpula-tion zone" beundaries?
4 Should there be a policy shif: in the cen-servative direction?
(I.e.,
shculd NRC in the future requi"e a grea:er degree of isolatien than that of the leas isclated
- of the already approved sites?)
With respect Oc each of these issues, we outline belcw wha FIRG o.ropcses> what scme other important options are, and what the staff describes as present practice and p~cpeses to include in the response.
Where we have cur own leanings at this time, we identify them.
We also include scme brief
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1.
Codification PIRG asks for codification via rulemaking.
Alternatives for policy articulation include a policy state-ment or continued use of staff documents, such as Regulatcry Guides and Standard Review Plans.
The staff agrees in principle that pclicy should be ccdified.
Mcwever, the staff would limit rulemaking a~t this time to a narrcw area (viz., amendment of 10 CFF 51, the enviren-mental "egulation, to call for " residual risk" evaluations when a guideline population density level is exceeded).
It would lean heavily en staff documents, which would lack i
the force of law.
We incline to broader rulemaking, because of the significance of the issues and the general recognition that the present siting regulation (10 CFR 100) is vague and cuite out-cf-date.
I 2.
Conceot of Exclusionarv Limits 4
?IRG wants to prohibit siting in places whe" surrounding population levels exceed certain stated numerical limits.
The staff defends a more flexible approach based on present t
practice, in which numerical populatien-level limits merely trigger nore rigorous examination, viz., cc=parison of estimated Class 9 accident consequences.
Exclusionary limits as proposed by ?IRG wculd be consisten:
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to siting in relatively iscla:ed areas.
The staff's flexible a,rproach would be acre consistent with a pclicy of viewing pcpulation levels as having an impcr:ance ec= parable to that of a number of other factors.
An internediate option would be to establish firm, exclu-sicnary upper limits at a relatively high level, introduce a favorable presumption at a substantially icwer level, and apply scaething like the present tradeoff ap.rcach to intermediate levels of population density
'ie are skeptical abcut the Class 9 acciden: cc secuence
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3 Parameters on Which Limits Should Se Placed
?IRG would specify:
. aximum allcwable pcpulation density (a)
M cut to a specified dis tance.
(b)
Minimum distance to exclusien area bcundar.v.
(c)
Minimum distance to low-pcpulaticn-cene boundary.
The staff agrees with parameter (a) (though with different numbers and only in a guideline sense).
Rowever, it r ej ect.=
minimum exclusion and L?Z distances, in crder tc preserve flexibility for tradeoffs, notably with "espect to ccm-pensating (containment) design features.
Alternatives to simple poculation density limits (e c..,
500 people per scuare mile cut to 30 miles) include con-sideration of locations of populaticn : enters above certain sizes, and weighting pcpulation according to proximity to the reactcr (such as the ccncent of " site.cc.rulation factor").
'.'e see merit in both minimum exclusion and L?Z distances.
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alternatives, and perhaps scae AC.:.S input.
1.~e ncte in this connection that, in a letter tc Mr. Gossick dated December 10, 1975, the Chairman of the ACES'recom-mended that thirteen specific items be considered as a means of imprcving site evluation, including the "desi ability of
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Cecree of Ccnservativeness
.'.n overarching issue is the degree of site isolation that the Cc= mission seeks to achieve.
Your views en this issue should be the main determinant in the assignment of specific numerical values to whatever parameters are selected fcr codificaticn as well as in decisions as to firmness or flexibility of limits.
?IRG's proposal would work a sucstantial shift of site approval practice in the ccnservative direction, partly I
because upper limits would beccme firm and exclusicnary, and partly because the specific numerical limits propcsed are more stringent than the present flexible guideline numbers (400. rec.cle.oer sc.uare mile cut to h0 niles vs. the staff's 500 out to 30).
The staff's flexible apprcach would allow future approval of sites ccmparable to the least isolated of existing sites.
The approach could also permit a liberal enough interpre-tation to lead to approval of substantially less isolated sites than any accroved in the cast, but we doubt that the policy would ce so liberally interpreted.
C.
SPECIFIC CO.'.:MENTS ON THE ?RCPCSED RESPONSE (Enclosure H)
Deenchasis cf ?coulation Factors Eoth the repcsed regulation Emendment and its supportin-e s
rationale tend to downplay the im-ortance of population e
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y importanu adverse factor.
In fact, the amendment's last sentence, by its uncualified ".Mothing shall prohibit..." phrasing, may tend to weaken any adverse presumption.
(p. 11!)
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s S~.C should be able new to specify which factors are accept-able and which are unacceptable in licensing plant sites.
The notion that every site is cotentially acceptable is without practical merit and should, in our judg:ent, be stricken from the rule.
Facters in Tradeoffs Agains: ?ctulatien Density
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a d en sit.y are alluded to in both the a endmen and the state-fy
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4.
. Alternative Site Evaluation olicy Tne proposed amendment appears to trea; pcpulation levels
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We agree that this can be a significan: aspect of the issue.
This aspect, however, should n)t be addressed in isolation
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5 "odificaticn vs.
. ejection of Petitioners' ?"crcsals "n scre respects the statement of considerations attacks the petitioners' propcsed formula frca what we believe is an unduly : arrow viewpoint.
70r ex u.ple:
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e The Ccrcissicn Januar'.
, 1979 maj cr pcpulation cente"s ).
(The sc:icn c." pcpula-ticn distribution also underlies the middle psra-graph on page 9 Ecwever, we dcubt the assertion there that the petiticners' numerical limits "' ould probably not change the relative distances between nuclear pcwer reactor sites and cetropolitan areas.")
6.
Emercencv ?lanning Considerr' ions In the denial of FIEG's.cro.cesed
_..arn distances (on.c.
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minimum icw-pc;ulation-:cne dis _.:ce prcpcsed 'cy the peti-tieners could 'or some sites " create an area which is too
_arge for effective protective actions."
This und'scussed assertien and its 2:plications need tc be reviewed, in our C; inion, in conjunction with:
(a)
Current rulemaking efforts on emergency planning requirements beyond the L.:Z (as presently d efined ).
(The staff is currently evaluating public ccmments on a proposed rule on this subject.)
(b)
The recommendations of the j oint NRC-EPA task force on emer eney c.lanninc.
(Those reccamen-s dations include the concept of an emergency planning rene, of about ten miles radius.)
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It dces not, hcwever, have an adequate information base for decision en them at this time.
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1.
That the Cc= mission is actively engaged in rethinking Part 100 in general, alternative site evaluation policy, and emergency plan-ning policy, all of which are germane to the a
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The Cc..iss'en Jar;Ery 4, 1979 Attached is a draft recuest for the Task Force's input.
. lease let us have your ccncurrence or comments by c.o.b.
January 9 Attachment as stated e
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f_g" "'4A h.._.iCLEAR REGULATORY COMMISSlh..
4 UNITED STATES i
(pglg WASHINGTON, D. C. 20555
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c January 5, 1979
+
OFFICE OF THE SECRETARY MEMORANDUM FOR: Lee V. Gossick, Executive ector for Operations 1
- FROM:
Samuel J. Chilk, Secretar s/:7
SUBJECT:
REQUESTFORPROGRES5 BRIE (IkhBYTHESITINGPOLICY TASK FORCE The Commission requests a progress briefing by the Siting Policy Task Force established pursuant to the Secretary's memorandum of August 8, 1978.
It'is desired that the briefing include but not necessarily be limited to the following topics:
1.
Work status, problems, and plans, including current thoughts on the scope of the Task Force's final report, scheduled for May 1979.
2.
Principal issues receiving Task Force attention and outline of current thinking on those issues.
3.
Ccmments on the possible role of the following concepts in a general policy framework for site evaluation:
(a)
Firm, exclusionary limits--for example, with respect to
~
upper beunds an nearby population levels or on seismic characteristics.
(b) Limits that, though not exclusionary, would trigger a presumption of " obvious inferiority" in site comparisons.
(c) The threshold concept; i.e., whetner some relatively in-significant impacts may be properly neglected in site.
evaluations.
4 Identification of any questions en which the Task Force seeks Commission guidance at this time.
The Ccomission wishes the briefing to take place before the end of January.
cc: Chairman Hendrie James Kelley Commissioner Gilinsky Ken Pedersen Commissioner Kennedy Daniel Muller Ccmmissioner Bradford Cc=missioner Ahearne Reference 3
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January 9, 1979
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V.EEDP.ANDUti FOR:
K. Pedersen, OPE (Tf"'(.17.YMh.QY J. Kelley, Act ng GC y
FROM:.
John Ahearne 5 f
~
SUBJECT:
. COPP.ENTS' ON'SECYi78-624' - 'PIRG PETITI0t! FOR RULEMAXING ON POPULATION DENSITY AROUND REACTOR SITES With respect to your recommendations for a Task Force briefing and no action regarding 78-624, I agree that we should not respond to the petition until further information is received and I agree with the proposed request for a b'riefing from the Task Force.
i I also believe it would be useful for the Task Force to be aware of any specific questions regarding the petition or the proposed responses that any Co.T.T.issioner may.have.
I am particularly interested in the
- following:
What anaTysi's can be presented to support any specific numerical criterion?
Nhat~ are the argumerrts-f6r treatihg transient popuTation by weighting the time in the area versus the petition's proposed use of the maximum transient population?
Whht are the arguments for using population density rather than p. cpu.1a.'.ica csaters?.
cc:
Chairman Hendrie Commissioner Gilihsky
- Cc 7.issicher' Kennedy Cc=.issioner Bradford Mr. Chilk y Reference 4
' +j. s * * " C0g Uf!!TEC STATES NUCLE AR REGULATORY CO'."M!SSION f..Ik - (6 Q
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20555 5,,.; ~T-&
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- ~ ~Y N 1 ] 1979
's o.....-
o7 FICE oF THE SEC 4 E TARY MEM0PANDUM FOR: Lee V. Gossick Executive Direc' - for Operations FROM:
Samuel J. Chil Secretary j
f i
SUBJECT:
COM' TENTS ON SECY-8-624 -- PIRG PETITION FOR RULEMAKING ON P0FULATION DENSITY AROUND REACTOR SITES Commissioner Ahearne has expressed his belief that it would be useful for the Siting Policy Task Force to be aware of any s'peci fic questions regarding the petition or the proposed responses that any Commissicner may have.
Commissioner Ahearne is particularly interested in the following:
-- What analysis can be presented to support any specific numerical cri te rion ?
-- What are the arguments for treating transient population by weighting the time in the area versus the petition's proposed use o f the maximum transient population?
-- What are the arguments for using population density rather than population centers?
cc: Chairman Hendrie Commissioner Gilinsky Co.1.missioner Kennedy Commissioner Bradford Commissionar Ahearne James Kelley, OGC Ken Federsen, OPE Caniel Muller, NRR KGoller, SD (%
coward Shapar, ELD e
Reference E
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MEMORANDUM FOI::
Chairman Hendrie Commissioner Gilinsky Commissioner Kennedy Co:mnissioner Bradford Commissioner Ahearne FROM:
Howard K. Shapar Executive Legal Director THRU:
Lee V. Gossick Executive Director fe?;staficiiF
SUBJECT:
SECY-78-624 -- PIRG PETITION FOR RULEMAKING ON POPULATION DENSITY AROUND REACTOR SITES i
Last week OELD received a copy of the joint OPE /0GC Commission memorandum on the PIRG petition for rulemaking cn poculation density around pcwer reactor sites.
This Office worked with the SD and NRR staffs on the pro-posed Commission response set forth in SECY-78-624, was aware that OGC and OPE had some problems with the recommendation, and understood that we would meet informally with OGC and OPE to discuss our different per-spectives on the issues.
That meeting never came about, so I think that it would be useful at this point to provide the Commission with our views on this matter.
We originally had some reservations about the Staff's proposed response to the PIRG petition but, after careful review, concluded that there was rational basis for the stated Staff belief that population density it; elf should not be a controlling consideration over other population distribution and safety and environceatal siting considerations, and concluded that the response met all legal requirements.
The OGC/0PE memorandum does not appear to raise any legal objections to the Staff proposal.
This Office has certainly not been one. of the defenders of Part 100 as presently written, and on several occasions we have urged Commissicn action on this anticuated regulation.
But the proposed response in SECY-78-624 is not based on seme~ presumed adequacy of present Part 100.
In deed,
the response recognizes that a broad re-examination of siting regulations is underway by the Commission.
The problem, as we saw it, was that the
Contact:
M. G. Malsch 492-7203 7eference 6
o' e*
. ongoing r2-examination of siting regulations would probably not result in new effective regulations for quite some time.
In the meantime, the Commission had an ineffective and ar,tiquated Part 100 as its only siting regulation, a staff siting review practice that seemed to be effective but which had no firm legal support, and a petition for rulemaking on siting matters that had been pending for some 21/2 years.
Beyond this, we saw some substantial advantages to ccdifying current Staff practice with regard to high-population density sites into the regulations--not the least of which was that the Staff practice as set forth in the Standard Review Plan was arguably at odds with more authoritative Commission adjudicatory decisions.
By separate memorandum frcm Mr. Chilk to Mr. Gossick, dated January 5, 1979, the Commission had already requested the Staff study group to look into issues raised by the PIRG petition, so in a sense the OGC/0PE recom-mandation has been adopted before SECY-78-c24 was even considered by the Commission.
L'e can't coce up with any overwhelming argument against more study of the PIRG petition by the Staff study group.
But, in the meantime, the Commission should recognize that it has virtually no effective limitations in its regulations on siting near populated areas.
/p ff'
(,
9 ward K. Shapar Executive Legal Director cc:
K. S. Pedersen, PE J. L. Kelley, OGC
~
R. B. Minogue, SD4 3
H. R. Denton, NRR w
/
e
t ENCLOSURE C Proocsed Regulatory Guidance on Accident Evaluations for Siting and Licensing Purcoses Enclosed are draft regulatory guidance on accident evaluations for siting and licgpatng purposes.
These proposed regulatory guides are i
intended to separate the assumptions and accident evauations to be per-formed for purposes of siting as required by 10 CFR Part 100 and for purposes of licensing as required by 10 CFR Parts 50 and 51.
The latter regulatory guide would combine and make consistent the evaluations per-formed for assessing the risk to the public health and safety and to the environment from design basis accidents for licensing nuclear pcwer plants.
Also enclosed is a draft regulatory guide (R.G.l.XXX) on atmospheric dispersion models for potential consequence assessraents to be used with the enclosed R.G.l.3 and 1.4.
This guide would require additional revisions to implement the enclosed design basis accident assumption guidance.
As stated in Enclosures "A" and "B", further changes would be required to other regulatory guidance given in the specific regulatory guides dis-cussed in Enclosure "B".
The draft meteorological model would replace the individual guidance given in accident related regulatory guides and would provide uniform guidance for accident meteorological assumptions.
Enclosure "C"
^
Prooosed Revisions to REG 0-uide 1.3 & l.a - Enclosure C to Procosed Revision of SECY 78-111
REGULATORY GUIDE 1.3 ASSUMPTIONS USED FOR EVALUATING LIGHT WATER POWER REACTOR SITING DISTANCE 5 FOR PURPOSES OF 10 CFR PART 100 A.
INTRODUCTION Section 100.ll(a) of 10 CFR Part 100 requires that each applicant should determine an exclusion area, a low population zone '.LPZ) and a population center distance of at least one and one-third times the distance to the outer boundary of the LPZ. The fission product release to be assumed for determining these distances should be based upon a major accident that would result in potential consequences not exceeded by those from any credible accident.
In the past, this major accident has been assumed to be the loss-of-c.colant accident resulting from the double ended rupture of a primary coolant pipe.
An additional requirement noted in 10 CFR Part 100 is that a release of appreciable quantities of fission products should be hypothesized for the evaluation of the above siting distances.
For current light water cooled nuclear power reactor designs, an identifiable reactor accident scenario can not be specified for meetino the assumptions of a major accident as noted in 10 CFR Part 100 for evaluating the suitability of a site.
Therefore, as noted in 10 CFR Part 100, the staff has hypothesized a fission product release from the reactor fuel in the care, which would not be exceeded frem any accident considered credib'?, for purposes of a 1.3-1 Enclosure "C"
site suitability analysis.
This guide gives acceptable assumptions that may be used for evaluating the siting distances def-:ned in 10 CFR Part 100.
In some cases, unusual site characteristics, plant design features or other factors may require different assumptions which will be con-sidered on an individual case basis.
The Advisory Ccmmittee on Reactor Safeguards has been consulted concerning this guide and has ccncurred in the regulatory position.
B.
DISCUSSION The NRC's nuclear power reactor site criteria require a judgment to be made during the regulatory process as to the proper balance between engineered safety features of a nuclear power facility and other means of assuring safety, including distance from large centers of population.
At the same time, the NRC has recognized that there is a need to take into account new data and improved safety devices as they are developed.
One means of accomplishing this objective is to use an accident analysis methodology whose implementation assures a consistent basis for judging site suitability, but still affords some flexibility of application.
For a number of years, the staff has utilized the evaluation of the radiological consequences of a postulated LOCA and an assumed large fission product release as one of the principal vehicles used to aid the review of.a proposed facility / site combination (population density and distribution, and feasibility of-protective measures on behalf of people in the immediate vicinity of the facility are other major considerations).
1.3-2 Enclosure "C"
~
While this practice originated about t-enty years ago, during a time when a core melt was considered a credible outcome of a LOCA, its continuance has stemmed from a view that the design bases for containment and related fissionproductremovalsystems[O$$the"last-ditch"protectionagainst ndoJL the release of appreciable quantities of radicactive materia 1 mitigate 4
releases of radioactive material greater than the ECCS design basis permits.
This overlap in protection was thought, particularly during the time when ECCS criteria were in a state of development, to provide an appropriate and prudent safety margin against unpredicted events in the course of accidents.
Recent analyses continue to confirm a large degree of margin associated with the source term used in the site evaluation process.
This guide lists acceptable assumptions that may be used in the determination of suitable site distances for purposes of Section 100.11 of 10 CFR Part 100.
It is recognized that certain of these assumptions may be regarded as physically unrealistic or excessively conservative for any predictable credible accident.
On the other hand, there are certain assumptions (such as ignoring the contribution of radioactive material contained in the primary coolant to the release of radioactive material from the fuel in the core and using adult man rather than critial segment of population for dose assumption) which are less conservative than might be the case.
In both instances, the staff has selected assumptions and models which, used in c0mbination, form an evaluation model which provides a consistent basis for aiding in judgments of site suitability.
Proposals 1.3-3 Enclosure "C"
for adjustments in the assumptions listed be. low will be considered if supported by new data on the effectiveness of various engineered safety features.
Proposals for adjustments which simply provide more realistic assumptions will generally not be supported by the staff unless their impact on the continuity and consistency with past siting practice has been assessed.
As noted above, the radiological assumptions used for purposes of 10 CFR Part 100 have also been used in establishing the design basis of containment and related fission product removal systems (see Regulatory Guides 1.7, 1.52, 1.73, 1.89, 1.96 and 1.141).
The evaluation models and assumptions set forth below provide guidelines.for determining the radiological environment that should be considered in evaluating these systems.
C.
REGULATORY POSITION 1.
The assumptions related to the release of radioactive material from the reactor fuel are as follows:
a.
Thermal power levels used for determining the inventory of fission products in the reactor core and available for release should be taken a: the design full power.
b.
One hundred percent of the noble gas inventory.
c.
Fifty percent of the iodine inventory.
1.3-4 Enclosure "C"
d.
Twenty-five percent of the cesium inventory, ten percent of the strontium inventory and one percent of other products frem the core, but all of this material is retained in the primary coolart.
2.
The assumptions related ta the transport of radioactive material within and from the primary containment are as follows:
a.
For purposes of estimating releases associated with primary con-tainment leakage, the inventory shculd be immedie'aij :vailable for lerikage.
That is, the inventory should be consrdered airborne inside primary containment.
b.
The airborne inventory should be assumed to be homogeneously distri-buted within the free value of the primary containment excapt as follows:
(1) Where the volume enclosed by the primary containment consists of compartments whfcn have limited atmosoneric exchange capability the initial distribution of tle airborne, inventory should be apportioned between these compartments such that the calculated release of radioactivity frcm the primary containment is maximized.
1.3-5 Enclosure "C"
(2) Where a secondary containment is proposed and bypass leakage is apportioned between various compartments which have Ifmited exchange capability, the initial distribution of the airborne inventory should ce apportioned between these compartments such that the calculatea release to the environment is maximized.
c.
Reduction 'in the amount of iodine available for leakage from the primary containment by plateout on internal containment surfaces may be taken into account, provided that suitably conservative models and assumptions are used.
Reference 1 provides one acceptable approach.
d.
Reduction in the amount of iodine available for leakage from the primary containment due to containment sprays may be taken into account, provided that suitably conservative models and assumo-tions are used.
Fcr purposes of the radiological assessment, the containment sprays may be assumed to operate at t
=0 provided that delivery of sprays commences within 90 seconds.
Reference 2 provides one acceptable approach.
e.
The reduction in the amount of iodine available for leakage from the primary containment by recirculating filter systems, ice condensers, suppression pools, or other engineered safety 1.3-6 Enclosure "C"
features may be taken into account, but the amount of reduction should be evaluated on an individual case basis.
f.
For purposes of evaluating the amount, and rate, of removal of iodine by various engineered safety features, 95.5% of the released icdine should be assumed to be in the form of elemental iodine, 2.5% in the form of particulate iodine and 2% in the form of organic iodines.
Time dependent rates of formation of organic iodines may be taken into account.
Reference 3 provides one acceptable approach.
i g.
The evaluation of effectiveness of containment sprays should take into account areas not covered by spray drops.
Atmospheric transfer rates between sprayed and unsprayed regions due to natural convection currents and turbulence induced by action of the spray of 2 volume exchanges per hour of the unsprayed region are generally acceptsble, provided that no restrictions to atmospheric transfer exist.
Forced air ventilation systems, designed to operate in the postaccident environment, are assumed to circulate air at 50% of their design flow rate.
h.
In evaluating the overall magnituce of the release from primary containment, where several containment-related ESFs are provided to remove fission products, one single failure snall be assumed.
1.3-7 Enclosure "C"
That failure should be selected on the basis of maximizing the calculated release.
i.
The primary reactor containment should be assumed to leak at the leak rate incorporated or to be incorporated as a technical speci-fication requirement at peak accident pressure for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (but not less than 0.1%/ day). The leak rate for the remaining duration of'the accident may be determined based on an assumed relationship of leak rate calculated differential pressure.
Peak accident pressure is the maximum pressure defined in the technical specifications for the containment.
3.
The assumptions related to the assumed effectiveness of secondary containment are as follows:
a.
Leakage from containment should be considered to go directly to the environment during periods when the secondary containment does not have a negative pressure, except that no direct cut-leakage need be assumed in the first 60 seconds provided that
-1/4" H O is developed thereafter.
2 b.
The removal of iodine by ESF filters may be accounted for, using the guidelines of Regulatory Guide 1.52.
1.3-8 Enclosure "C"
c.
Primary containment leakage should be assumed to be transferred directly to the exhaust with no mixing or holdup in the region being filterei (exceptions will be reviewed on an individual case basis).
d.
For systems involving recirculation of exhaust air back into secondary containment, 50% mixing is acceptable.
Greater values may be considered depending on header placement.
e.
Bypass leakage should be included. Where such' leakage is through water, credit for retention of iodine may be taken into account.
4.
Acceptable assumptions for estimating the radiological consequences of released radioactivity are:
a.
The exclusion boundary distances should be computed based on the distance from the center of the primary containment to the center of eaca of 16 ordinal points along the exclusion area boundary as defined in 10 CFR Part 100, Section 100.3(a).
However, where releases occur from known locations such as the exhaust from standby gas treatment systems, that location should be used.
1.3-9 Enclosure "C"
b.
Meteorological dispersion conditions for releases from primary and secondary containment to the environment should be as deter-mined from Regulatory Guide 1.XXX.
c.-
No correction should be made for depletion of the effluent plume of radioactive iodine due to deposition on the ground.
d.
For the exclusion area distance, the breathing rate of a working 4
adult for the two hour exposure should be assumed to be 7 x 10 m /second.
9 e.
For the outer boundary of the LPZ, the breathing rate of an adult for the first 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> exposure should be assumed to be
~4 3
3.47 x 10 m /second and for the 8 to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> exposure,
~4 3
1.75 x 10 m / seconds.
If significant containment leakage or radiation exposure can continue for time periods greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the breathing rate should be assumed to be 2.32 x
~4 3
10 m /second.
f.
Oose conversion factors should be as given in Regulator / Guice 1.109 for the adult (Table E-7).
g.
External whole body doses should be calculated using the
" Semi-Infinite Cloud" assumptions or Finite Cloud assumotions.
If finite cloud assumptions are used, the meteorological 1.3-10 Enclosure "C"
parameters used should be consistent with the general acce..ance criteria of Regulatory Guide 1.XXX (the worst 5% dispersion conditicns for whole body immersion dose may be different from that for inhalation dose).
An analysis should be performed to determine what combination of dispersion conditions and gamma ray travel (see Appendix F of R.G. 1.109) corresponds to the worst 5% dose.
Beta dose need not be considered in evaluating whole body dose.
h.
The average gamma energies emitted per disintegration as given in the Table of Isotopes are acceptable (Reference 4).
i.
The computed doses should be compared against the guideline values of 10 CFR Part 100.
Should there exist significant uncertainties regarding tne assumed effectiveness of the a
proposedcontainmen)/eiatedengineeredsafetyfeaturesand/or meteorologir.al data, the possible increase in the necess "j siting distance for the exclusion area and the outer boundary for the LPZ should be determined and a proposed contingency resolution stated to meet 10 CFR Dart 100 requirements.
5.
For purposes of evaluating the design bases for the engineered safety features employed for purposes of mitigating the radiological con-sequences of a LOCA, the time-decendent concentration in containment sumps on building surfaces and in builcing atmospheres snould :e 1.3-11 Encic?are "C"
calculated using the assumption in Positions C.1-C.4 above.
(Regulatory Guides 1.7, 1.52, 1.73, 1.89, 1.96, 1.97, and 1.141 identify certain equipment and analyses where the radiological environment set forth herein is to be used).
D.
IMPLEMENTATION Except in those cases in which the applicant proposes an acceptable alternative msthod for complying with specified portions of the Commission regulations, th* methods described herein will be used in the evaluation of applicatiot', docketed after (to be filled in at time of issuance of the guide).
E.
REFERENCES 1.
R. Sherry, " Computation of Time Dependent Plateout Following a Design Basis LOCA (August 1978).
2.
A. K. Postma and P. S. Tam, " Technological Bases for Models of Spray Washout of Airborne Contaminants in Containment Vessels," NUREG-CR-0009 (August 1978).
3.
A. K. Postma and R. W. Zavadoski, " Review of Organic Iodide Formation Under Accident Conditions in Water-Cooled Reactors," WASH-1233, October 1972.
4 C. M. Lederer et al., Table of Isotoces, Sixth Edition, University of California, Berkeley, Lawrence Radiation Lacoratory.
1.3-12 Enclosure "C"
REGULATORY GUIDE 1.4 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES TO THE PUBLIC FRCM DESIGN BASIS ACCIDENTS IN LIGHT WATER POWER REACTORS FOR PURPOSES OF 10 CFR PARTS 50 AND 51 A.
INTRODUCTION Sections 50.35(c) and 50.56 of 10 CFR Part 50 require that the Com-mission not authorize operation a f the facility until it has found that the final design provides reason ible assurance that the health and safety of the public will not be endangtred by operation of the facility in the manner proposed.
Sections 51.22 s,d 51.26 of 10 CFR Part 51 require that the Director of Nuclear Reactor Regu?ation prepare a draft environmental impact statement and a final environmental impact statement for light-water-cooled nuclear power reactors.
The potential radiological conse-quences to the public from postulated design basis accidents for the final facility design on the specified site should be evaluated using realistic assumptions.
This guide gives acceptable assumptions that may be used to assess the potential radiological consequences to the puolic from design basis accidents in light water power reactors.
The assumptions for release of radioactive material from fuel with clad failure are related to the 1.4-1 Enclosure "C"
postulated loss-of-coolant accident conditions necessary to meet Appen-dix K, 10 CFR Part 50 requirements.
The site specific distances estab-lished in accordance with 10 CFR Part 100 from assumptions given in Regulatory Guide 1.3 will be used in the realistic evaluation.
In scme cases, unusual site characteristics, plant design features or other factors may require different assumptions which will be considered on an individual case basis.
The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regula-tory postion.
B.
DISCUSSION For a number of years, the staff has utilized the evaluation of the radiological consequences of postulated design basis accidents as one of the principal means to review a proposed facility / site comoination.
For the LOCA, an assumed large fission product release has been used by the staff for more than twenty years to establish site suitability.
During the time period when ECCS criteria were being developed, the fission product release for the LOCA was greater than the ECCS design basis would permit.
This overlap was provided as an appropriate and prudent safety margin against unpredicted events for the design of the containment and related fission product removal systems.
Meanwhile, the staff utilized different assumpticas for the LOCA and other design basis accidents to evaluate the environmental impact which used a small fraction of the tission product release from the LOCA assumed for the site suitaoility analysis.
These assumptions were given in a proposed annex to Accendix 0, 10 CFR Part 50 and are given as Appendix I to Regulatory Guice 1.2.
1.4-2 Enclosure "C"
This guide lists acceptable assumotions that may be used in the determination of realistic radiological consequences from potential design basis accidents in light water power reactors for purposes of 10 CFR Parts 50 and 51. Only the assumptions for release of radicactive material from fuel with clad failure (gap activity) related to the ECCS design basis is given for the design basis LOCA.
Other design basis accidents should use radioactive material releases specified by other regulatory guides or related to speci-fic activity levels stated in the technical specifications.
A 1[.REGULATORYPOSITI0hl.
The assumptions related to the release of radioactive material from the reactor fuel (gap activity) for a LOCA under ECCS design basis are as follows:
a.
Thermal power levels used for determining the inventory of j
fission products in the reactor core and available for release should be taken as the design fuil cower.
b.
Six percent of the noble gas inventcry.
I c.
Three percent of the iodine inventory.
2.
The assumotions related to the transport of radioactive matarial witnin and from the primary containment are as follows:
1.4-3 Enclosure "C"
For purposes of estimating releases associated with primary con-a.
tainment leakage, the inventory should be immediately available for leakage. That is, the inventory should be considered airborne inside primary containment.
5 b.
The airborne inventory should be assumed to be homogeneously distri-buted within the free volume of the primary containment except as follows:
(1) Where the volume enclosed by the p.r.3ay;y contairaent consists
~
of compartments which have limited atmospheric exchange capability the initial distribution of tne airborne inventory should be apportioned between these compartments such that the calculated release of radioactivity from the primary containment is maximized.
6*} Where a secondary containment is proposed and bypass leakage is apportioned between various compartments which have limited exchange capability, the initial distribution of the airborne inventory should be apportioned between these compartments.
For purposes of estimating releases prior to containment isolation c.
for facilities where containment purging is routinely practiced, 1.4-4 Enclosure "C"
only the activity associated with pre-accident coolant contami-nation should be assumed to be available for leakage, provided that isolation is effected promptly.
5 d.
Reduction in the amount of iodine available for leakage from the primary containment by plateout on internal containment surfaces may be taken into account, provided that suitably conservative
.j models and assumptions are used.
Reference 1 provides one acceptable approach.
e.
Reduction in the amount of iodine available for leakage from the primary containment sprays may be taken into account, provided s
that suitably' conservative models and assumptions are used.
For purposes of the radiological assessment, the containment sprays may be assumed to operate at t = 0 provided that delivery of sprays commences within 90 seconds.
Reference 2 provides one acceptable approach.
f.
The reduction in the amount of iodine available for leakage from the primary cortainment by recirculating filter systems, ice condensers, suppression pools, or other engineered safety features may be taken into account, but the amount of reduction should be evaluated on an individual case basis.
i 1.4-5 Enclosure "C"
g.
For purposes of evaluating the amount, and rate, of removal of iodine by various engineered safety features 95.5% of the released iodine should be assumed to be in the form of elemental iodine, 2.5% in the form of particulate iodine and 2% in the form of organic lodines.
Time dependent rates of formation c organic fadines may be taken into account.
Reference 3 provides one acceptable approach.
h.
The evaluation of effectiveness of containment sprays should take into account areas not covered by spray drops.
Atmosp!t'ic trcnsfer rates between sprayed and unsprayed regions due to natural convection currents and turbulence induced by action of the spray of 2 volume exchanges per hour of the unsprayed region are generally acceptable, provided that no restrictions to atmospheric transfer exist.
Forced air ventilation systems, designed to operate in the postaccident environment, are assumed to circulate air at their tested flow rate.
i.
The primary reactor conta.inment should be assumed to leak at the leak rate requirements stated in the technical specifications for the pressure ar.d temperature conditions predicted by the ECCS design basis conditons for the course of the LOCA.
The leak rate should follow the calculated differential pressure for the design basis accident being evaluated.
1.4-6 Enclosure "C"
3.
The assumptions related to the assumed effectiveness of secondary containment are as follows:
a.
Leakage from containment should be considered to go directly to the environment during periods when the secondary containment does not have a negative pressure, except that no direct out-leakage need be assumed in the first 60 seconds provided that
-1/4" H O is developed thereafter.
2 b.
The removal of iodine by ESF filters may be accounted for using the test requirements given in the technical specifications.
c.
Primary containment leakage should be assumed to be transferred directly to the exhaust with credit for mixing or holdup in the region being filtered.
d.
For systems involving recirculation of exhaust air back into secondary containment, 50% mixing is acceptable. Greater values may be considered depending on header placement.
e.
Bypass leakage should be included. Where such leakage is through water, credit for retention of iodine may be taken into account.
4 Acceptable assumptions for estimating the radiological consecuences of released radioactivity are:
1.4-7 Enclosure "C"
a.
The exclusion boundary distances and outer boundary distances of the LPZ should be those distances determined for purposes of 10 CFR Part 100 (Regulatory Guide 1.3),
b.
Meteorological dispersion conditions for releases from primary and secondary containment to the environment should be as deter-mined from Regulatory Guide 1.XXX.
Releases associated with containment purge and/or ECCS leakage that may occur subsequent to the accident should be computed and added to the LPZ dose associated with containment leakage, assuming meteorology con 2
1 sistent with 1-4 day estimates of Regulatory Guide 1.XXX.
c.
No correction should be made for depletion of the effluent plume of radioactive iodine due to deposition on the ground.
d:
For the exclusion area distance, the breathing rate of a working adult for the two hour exposure should be assumed to be 7 x 10 3
m /second.
e.
For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of the most critical
-5 individual should be assumed, i.e., children, to be 7.3 x 10
' cubic meters per second.
From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident the breathing rate for the same critical individual,
-5 i.e., children, should be assumed to be 3.1 x 10 cubic meters per second.
For postulated cesign casis accidents in wnicn 1.4-8 Enclosure "C"
signifiant containment leakage could contine for time periods greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the breathing rate should be assumed to S
be 4.4 x IO cubic meters per second for the child.
These breathing rates are to be used for an individual within and beyond the established low population zone.
f.
Dose conversion factors should be s given in Regulatory Guids 1.109 (Tables E-7 and E-9) g.
External whole body doses should be calculated using the
" Semi-Infinite Cloud" assumptions or Finite Cloud assumptions.
If finite cloud assumptions are used, the meteorological para-meters used should be consistent with the general acceptance i
criteria of Regulatory Guide 1.XXX (the worst 25% dispersion conditions for whole body immersion dose may be different frem that for inhalation dose).
An analysis should be performed to i
determine what combination of dispersion conditions and gamma ray travel (see Appendix F of R.G.1.109) corresponds to the worst 25% dose.
Beta dose need not be considered in evaluating whole body dose.
h.
The average gamma energies emitted per disintegration as given in the Table of Isotopes are awceptable (Reference 4).
1.4-9 Enclosure "C"
i.
Twenty-five percent meteorology for the region surrounding the site out to 50 kms should be used in the co.. sequence analysis.
j.
Population exposures for critical organs should be determined for the population distribution predicted at the time of plant startup for predicted dose levels greater than 5 mrem within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for airborne pathways and within 1 year for ingestion pathways.
The integrated population exposures in man rem per design basis accident scenario for each critical organ should not include potential exposures L an individual (or critical organ) below 5 mrem.
The average breathing rate of an adult of
~4 3
2.6 x 10 m /sec should be used for determining population exposures.
D.
IMPLEMENTATION Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission regulations, the methods described herein will be used in the evaluation of applications docketed after (to be filled in at time of issuance of the guide).
E.
REFERENCES 1.
R. Sherry, " Computation of Time Dependent Plateout Following a Design Basis LOCA"(August 1978).
1.4-10 Enclosure "C"
2.
A. K. Postca and P. S. Tam, " Technological Bases for Models of Spray
- Washout of Airborne Contaminants in Containment Vessels," NUREG-CR-0009 (August 1978).
3.
A. K. Postma and R. W. Zavadoski, " Review of Organic Iodide Formation Under Accident Conditions in Water-Cooled Reactors," WASH-1233, October 1972.
4.
C. M. Lederer et al., Table of Isotooes, Sixth Edition, University of California, Berkeley, Lawrence Radiation Laboratory.
4 i
4 1.4-11 Enclosure "C"
Enclosure E REGULATIONS FOR SITING OF NUCLEAR FACILITITIES -
REVISED 10 CFR PART 100 Since their promulgation by the Atomic Energy Commission (AEC) in April 1962, the reactor site criteria set forth in 10 CFR Part 100 have served as the framework for evaluation of proposed sites for stationary power and test reactors from the standpoint of protection of the health
'and safety of the public. As the Statement of Consideration which accompanied publication of the effective Part 100 in the Federal Register (27 FR 3509) indicated, the effective Part was intended to reflect the AEC siting practices being used at that time.
Following its organization under the Energy Reorganization Act of
~
1974 (Public Law 93-438), the Nuclear Regulatory Commission (NRC) stated 9
its. intention of reviewing those of its regulations and procedures pertaining to the licensing and regulation of nuclear facilities and materials which were originally promulgated by the AEC, with a view to considering what changes should be made. As the Statement of Considera-tion which acccmpanied publication of an amendment to Part 100 in the Federal Register (40 FR 26525) indicated, a general examination of power reactor siting regulations and policies was underway as a separate matter. As part of the above stated effort, the 50 staff has prepared the attached revision to 10 CFR Part 100, which would extend the siting regulations to include all nuclear facilities licensed pursuant to 10 CFR Parts 50 and 51. The SD staff considers that the revised siting regulations should encompass all factors which impact on siting decisions 1
Enclosure "E" Drocosed Devision to in cro eart 100 - Enclosure E to ECY 77-22S
elsewhere in Title 10 rules and regulations. The 50 staff considers the re-evaluation of these siting regulations to be especially timely in view of concerns that have been recently expressed by public officials at various levels of government, public interest groups and various components of the nuclear industry.
The regulations covering power and test reactor siting have been re-evaluated from the standpoint of expanding the regulations to include all nuclear facilities licensed pursuant to 10 CFR Part. 50 and 51 and to include siting factics other than those related to radiological risks to the health and safety of the public from the release of radioactive material due to postulated accidents from said nuclear facilities.
Particular interest is expressed in receiving views and comments on the following:
1.
The SD staff has prepared a draft version of the main body for a revised Part 100 which is intended as a basis for discussion and comment. This main document would establish the general rules and guiJance describing good practice for siting of any nuclear facility licensed pursuant to 10 CFR Parts 50 and 51. The appendices to the main document could contain: 1) specific criteria of design to be met by nuclear facility types such as light water cooled power reactors and 2) specific criteria for determining the site related phenomena, such as geologic and hydrologic features, to be acccm-modated by design. The current Appendix A to 10 CFR Part 100 with modifications would reflect such specific criteria. The appendices 2
Enclosure "E"
to the main document should contain the general statements of policy relating to siting evaluations and implementation.
2.
The SD staff contemplates the expansion of the regulations for nuclear facility siting pursuant to 10 CFR Part'100 to include all nuclear facilities licensed pursuant to 10 CFR Parts 50 and 51.
3.
The SD staff contemplates tne expansion of the regulation for nuclear facility siting pursuant to 10 CFR Part 100 to include standards and methodologies for site selection and for assessing environmentally related siting requirements.
3 Enclosure"E"
PART 100 SITING OF NUCLEAR FACILITIES GENERAL PROVISIONS Sec. 100.1 Purpose 100.2 Scope 100.3 Definitions 100.4 Communicatier.s 100.5 Interpretations EXEMPTIONS ANC REOUIREMENTS 100.10 Specific Exemptions 100.11 Additional Requirements SITING CRITERIA 100.20 General Criteria 100.21 Criteria for Design Basis External Natural Events 100.22 Criteria for Design Basis External Man-Induced Event:
100.23 Criteria for Defining Potential Effects of the Nuclear Facility on the Region 100.24 Criteria for Regional Distributien of Population EVALUATION PROCEDURES FOR EXTERNAL NATURAL EVENTS 100.30 Floods 100.31 Tsunami or Seiche 100.32 Seismic Events 100.33 Tornadoes or Hurricanes 4
Enclosure"E" s
e em-
Slope Instability 100.34 Site Surface Collapse and 100.35 Subsidence and Rebound Other Natural Pehnomena 100.36 EVALUATION PROCEDURES FOR EXTERNAL MAN-INDUCED EVENTS 100.40 Aircraft Crash 100.41 Chemical Explosions and Toxic Releases 100.42 Ultimate Heat Sink 100.43 Physical Security Other Man-Induced Events 100.44 EVALUATION PROCEDURES FOR ENVIRONMENTAL ASPECTS 100.50 Meteorology and Atmcspheric Dispersion 100.51 Hydrology and Water Dispersion 100.52 Population Distribution 100.53 Land and Water Uses 100.54 Baseline Radioactivtty 100.55 Baseline Ecological Conditions BACKFITTING Backfitting 100.60 ENFORCEMENT Violations 100.70 APPENDICES Appendix A - General Statement of Policy For Nuclear Reactor Site Evaluations 5
Enclosure"E"
GENERAL P".0 VISIONS I 100.1 Purpose The regulations in this part establish procedures and criteria for the siting of nuclear facilities licensed pursuant to the Atomic Energy Act of 1954 as amended and Parts 50 and 51 of this chapter. The accept-ability of a site is closely related to the design of the proposed facility.
Any applicant who belii!ves that factors other than those set forth in this part should be considered by the Cormnission will be expected to demon-strate the applicability and significance of such factors.
From the public health and safety and environmental protection point-of-views, a site is acceptable if technical solutions to site-related problems can give assurance that the proposed nuclear facility can be built and opera-ted within an acceptably low risk to the public and to the environs. The regulations in this part establish the objectives and minimum requirements which must be fulfilled to prov'fe adequate safety and environmental protection for the construction, operation and decommissioning of the proposed nuclear facility on the proposed site.
I 100.2 Scope The regulations cents:ned in this part apply to applicaticns filed under Part 50 of this chapter. The scope of the regulations contained in this part encompasses all site and site-facility interaction factors related to the proposed nuclear facility. These factors include, but are not limited by, postulated accidents, natural and man-made events external to the nuclear facility, normal operation of the nuclear facility and effects of the nuclear facility on the environment and those that 6
Enclosure "E"
can be important to the health and safety of the public. The procedures and criteria established in this part apply to nuclear fucilities of a general type and design on which experience has been developed but can be applied to any nuclear facility licensed in accordance with Parts 50 and 51 of this chapter. For nuclear facilities that are novel in design or d
unproven such as prototypes or pilot plants, these procedures and criteria should be applied in a. manner that takes into account the lack of experience.
The criteria set forth in this chapter are declarative in that the information to be considered is identified for implementing the general statement of policy for nuclear reactor site evaluations as given in Appendix A.
The specific guidelines and criteria approcriate to each l
general criterion set forth in this chapter wil? be included in the appen-dices to this chapter as they become generally applicable.
s 100.3 Definitions As used in this part:
(a)
"Act" means the Atomic Energy Act of 1954 including any amendments thereto.
(b) " Commission" means the Nuclear Regulatory Commmission or its duly authorized representatives.
(c) " Common defense and security" means the common defense and security of the United States.
7 Encl osure "E"
(d) " Controlled area" means that area immediately surrounding the facility, the use of which is controlled by the licensee during the period of a license and in which is permitted that operation encompassed by the facility license issued pursuant to Part 50 of this chapter.
(e) " Controls" when used with respect.to nuclear reactors means apparatus and mechanisms, the maninulation of which directly affects the reactivity or power level of the reactor.
" Controls" when used with respect to any other facility means apparatus and mechanisms, the manipu-lation of which could affect the chemical, physical, metallurgical, or nuclear process of the facility in such a manner as to affect the pro-tection of health and safety against radiation.
(f) " Design basis" means the parameter values associated with that level of severity of an external event or combination of events selected for design of all or any part of a nuclear facility to ensure that the structures, systems and components important to safety or enviro r. ental protection (in relation to that event or combination of events) will maintain their integrity and will not suffer loss of func tion during or after the event or before completing its design fanction. These values may be (1) restraints derived frem generally accepted " state of the art" practices for achieving functior:a1 goals, or (2) requirements derived from analysis (based on calculation and/or experiments) of the effects of a postulated event for which a structure, system. or component must meet its functional goals.
8 Enclosure "E"
(g) " Design basis for external events" means that (1) estimates of severe natural events are to be used for deriving design basis for which such estimates will be based on consideration of historical data of the associated parameters, physical data or analysis of upper limits of the physical processes involved; and (2) estimates of severe external man-induced events are tc be used for deriving design basis for which such estimates v411 be based on analysis of human activity in the region taking into account the site characteristics and the risks associated with the i
event.
(h) " Historical data" means a compilation of physical data con-cerning a particular type of event derived from human history.
(1) "NEPA" means the National Environmental Policy Act of 1969 including any amendments thereto.
(j) " Neighboring area" means that area immadiately surrounding the controlled area in which population distribution and density and land and water uses are considered with respect to the possibility of implementing contingency measures.
(k) " Nuclear reactor" means an apparatus, in which nuclear fission may be sustained in a self-supporting chain reaction at a controlled rate.
(1) " Person" means (1) any individual, corporation, partnership, firm, association, trust, estate, public or private institution, group, Government agency other than the Commission, any State or any political subdivision of, or any political entity within a State, any foreign gov-ernment or nation or any political subdivision or any such government or 4
9 Enclosure'"E"
- = =..
p
nation, or other entity; and (2) any legal successor, representative, agent, or agency of the foregoing.
(m) " Population" means all the critical organisms living in a given area. The critical organisms shall be those ecological and biota systems adversely affected by change,in conditions due to the construction and operation of the nuclear facility, including people in the area.
(n) " Region" means a geographical area surrounding and including the site sufficiently large to contain (1) all the features related to a phenomenon or to the effects from a particular event and (2) all measur-able effects of environmental impact due to the facility both radiological and non-radiological.
(o) " Site" means that area on which a nuclear facility may be located. The site includes the controlled area.
(p) " Ultimate heat sink" means the complex of coolant sources for a r.uclear reactor necessary to dissipate resideal heat after reactor shutdown or after an accident.
(q) " United States," when used in a geographical sense, includes all Territories and possessions of the United States, the Canal Zone, and Puerto Rico.
I 100.4 Communications Except where otherwise specified, all communications and reports concerning the regulations in this part, and applications filed under them should be addressed to the Executive Director for Operations, U.S. Nuclear 10 Enclosure"E"
Regulatory Commission, Washington, D.C. 20555.
Communications, reports and applications may be delivered in person at the Commission's offices at 1717 H Street N.W., Washington, D.C. or at 7920 Norfolk Avenue, Bethesda, Md.
s 100.5 Interpretations Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel will be recognized to be binding upon the Com-mission.
EXEMPTIONS AND REOUIREMENTS i 100.10 Specific Exemptions The Commission may, upon application by an interested person, or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will ot endanger life or property and are otherwise in the public interest.
n s 100.11 Additicnal Requirements The Commission' may, by rule, rey;ulation, or order, impose upon any licensee such requirements in addition to those established in the regu-lations in this part, as it deems appropriate or necessary to protect health and the environment and to minimize danger to life or property.
11 Enclosure "C"
SITING CRITERIA 5 100.20 General Criteria I
(a) Site characteristics which may directly affect the safety or environmental aspects of the nuclear facility shall be investigated and assessed.
(b) Proposed sites for nuclear facilities shall be exarnined with respect to the frequency and the severity of external natural and man-l induced events that could affect the safe operation of the nuclear facility.
(c) Design basis external events shall be determined for each combination of proposed site and proposed nuclear facility.
(d) Proposed sites with design basis external events for which ade-quate protection cannot be provided through nuclear facility design shall be deemed unsuitable for the location of a proposed nuclear facility.
(e) Site related design basis events shall be evaluated and reviewed before construction of the nuclear facility is approved.
(f) For each proposed site, the potential for radiological and non-radiological consequences in the region shall be evaluated with due consideration of the characteristics of the population, including its distribution.
(g) For each proposed site, the potential for environmental i pact m
to the' region shall be evaluated with due consideration of the character-istics of the regional environs, including its historical and aesthetic value.
12 Enclosure "E"
I 100.21 Criteria for Design Basis External Natural Events (a) Natural phenomena which may exist or can occur in the rcgion of a proposed site shall be identified and assessed according to their The potential effects on the safe operation of the nuclear facility.
important natural phenomena for which design bases should be derived shall be identified.
(b) Historical records of the occurrence and severity of those important natural phenomena shall be collected for the region and care-fully analyzed for reliability, accuracy and completeness.
(c) Appropriate methodologies shall be adopted for establishing the The' mcthodo-design basis natural events for important natural phenomena.
logies should be justified as being compatible with the characteristics of the region and the current state of knowledge.
5 100.22 Criteria for Design Basis External Man-Induced Events (a) The regio-shall be examined for man-made facilities and activi-ties that might endanger the proposed nuclear facility. The important man-induced phenomena for which design basis external man-induced events should be derived shall be identified.
(b) Information concerning the occurrences and severity of those important man-induced phenomena shall be collected and analysed for reliability, accuracy and completeness.
(c) Appropriate methodology shall be adopted for establishing the The method-design basis external man-induced events for those phenomena.
ology should be justified as being compatible with the characteristics of the region and the 1rrent state of knowledge.
13 Enclosure "E"
5 100.23 Criteria for Defining Potential Effects of the Nuclear Facility on the Region (a) Each site shall be examined with respect to the effects on people in the region resulting from the release of radioactive materials under normal and accident conditions; in this evaluation unusual regional and site characteristics shall be taken into account.
(b) Each site shall be examined with respect to the effects on the regional environment resulting from construction, operation and decom-missioning of the nuclear facility; in this evaluation unusual regional and site characteristics shall be taken into account.
(c) Effects which would otherwise be unacceptable shall be compen-sated for by the nuclear facility design or the site shall be deemed unsuitable.
I 100.24 Criteria for Regional Distribution of Population (a) The proposed site shall be studied to evaluate the present and future character and distribution of the htman population of the region.
Such a study, which should include evaluation of present and future uses of land and water within the region, shall also take into account any special characteristics which may influence the potential consequences of a release of radioactive material during the operational lifetime of the neclear facility.
(b) A controlled area and a neighboring arca shall be established for each site.
Erciosure "E" 14 F
(c) The size of the controlled area for a proposed site shall be defined by the authority of the licensee to determine all activities including exclusion or removal of personnel and property from the area.
(d) The neighboring area of a proposed site shall be evaluated from the perspective of the potential for adverse consequences to the human population or environment and of the capability of implementing protective measures as may be necessary to mitigate the immediate effects of a release of radioactive material.
(e) The distribution of the human population in the region sur-rounding the site shall be evaluated from the perspective of the potential for adverse consequences to regional populations from normal and potential accidental releases of radioactive material or other toxic materials and from construction, operation and decommissioning of the nuclear facility during its lifetime.
(f) Effects which would otherwise be unacceptable shall be compen-sated for by the nuclear fa-ility design or the site shall be deemed un, suitable.
EVALUATION PROCEDURES FOR EXTERNAL NATURAL EVENTS
$ 100.30 Floods (a) A suitable hydre. meteorological and hydrologic model shall be developed taking into consideration all relevant changes in characteris-tics of the region which (11 have occurred during the historical time period or (ii) are planned or may occur during the lifetime of the nuclear facility. The design flood caused by precipitation shall be derived from the model.
15 Enclosure "E"
(b) The design basis flood shall include the height of the water (including waves), the duration of the floods and the flow conditions.
(c) At coastal and similar sites the possibility of flooding by a combination of high tide, wind effects on bodies of water, and wave actions shall be examined and a design basis event shall be derived from the site.
(d) Information concerning upstream water control structures shall be analyzed to determine whether (i) such structures are likely to fail from seismic, flood, or other causes, or (ii) the design of the nuclear facility is able to accommodate the postulated failure of any upstream structure.
(e) If examination of the nuclear facility indicates that it cannot accommodate safely all the effects of the failure of tha upstream structure, the design basis for the nuclear facility shall be modified to include all such effects or the site shall be deemed unsuitable.
(f) The possibility of temporarily blocking rivers (e.g., landslides or ice) so as to cause floodfr.g and associated phenomena at the proposed site shall be examined.
5 100.31 Tsunami or Seiche (a) If tha potential for a tsunami or seiche exists, historical tsunami or seiche data of the shore region in which the site is located shall be collected and critically examined for reliability and relevance to the site.
16 Enclosure "E" e
(b) The possibility for tsunami or seiche to be generated by local off-shore seismic events shall be evaluated with respect to known seismic records and seismotectonic characteristics.
(c) On the basis of the above, information for the shore region and by comparison with similar shore regions, the frequency of occurrence, magnitude and height of regional tsunami or seiche shall be estimated and used to derive a design basis tsunami or seiche taking into account amplifica-tion phenomena due to the shore configuration at the site.
(d) The main parameters of the design basis tsunami or seiche, including drawdown, wave height and possible physical effects on the site, shall be derived from the above informatien.
5 100.32 Seismic Events (a) The seismology and the geology of the region of a proposed site shall be evaluated.
(b) In some regions where the seismic history is well known, the historical data provides an acceptable basis for characterising future locations and intensities of earthquakes. Such regions exhibit a correla-tion between earthquake activites and structures.
(c). An alternative method is the use of the seismotectonic method which characterizes regions according to their relative homogenity of geological structural features. Such regions represent the boundaries of seismic activity not correlated with structures.
(d) The design basis vibratory ground motion shall be expressed by appropriate parameters such as envelopes of frequency response spectra for 17 Enclosure "E"
various damping factors, duration of shaking, and time histories of the 8
acceleration.
(e) The design basis vibratory ground motion for offsite structures, the failure of which could affect the safe operation of the nuclear facility, shall be derived using the same method adopted for the nuclear facility and the effects on those structures shall be evaluated.
(f) An investigation for surface faulting shall be. made at the site and in its vicinity unless adequate evidence is available to indicate that surface faulting has not occurred in the region.
(g) Detected surface faults shall be examined to determine if they Those are capable of causing significant displacement or earthquakes.
capable faults which may influence the surface of the site shall be 4
investigated before the site is deemed suitable with respect to surface f aulting.
In evaluating surface faulting, the strength of the evidence shall be taken into consideration with respect to the extent of the investigation and the method to be used.
The potential for liquefaction of the subsurface of the proposed (h) site shall be evaluated using the 6 sign basis vibratory ground motion.
The evaluation shall include the use of accepted soil investigation and analytical methods with a margin of safety provided to take into account the uncertainties in the determinaion of soil characteristics and calcu-lational methods.
5 100.33 Tornadnes or Hurricanes (a) The occurrence of tornadoes or hurricanes in the region of interest shall be evaluated.
If tornadoes or hurricanes have occurred, detailed historical data shall be collected.
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(b)
If historical data are inadequate to determine a design basis tornado or hurricane, then they shall be supplemented with data from other regions for which tornado or hurricane statistics are available and which have similar climatological chracteristics.
(c) The design basis tornado shall be derived and expressed in terms of rotational wind speed, translational wind speed, radius of maximum rotational wind speed, pressure differentials, and rate of chanse of pressure.
(d) The design basis hurricane shall be derived and expressed in terms of extreme wind speed, radius to extreme wind speed, pressure differential, translational speed and storm track.
[
(e) Potential missiles associated with design basis tornadoes or hurricanes shall be evaluated and included in the formulation of the design basis.
~
5 100.34 Slope Instability (a) The vicinity of the site shall be evaluated to determine. the potential for slope instability (such as land and rock slides ar.d snow avalanches) which could affect the safety of the nuclear facility.
(b). If such potential exists, a detailed study shall be made which 1
includes consideration of vibratory ground motion derived by the method used for establishing the design basis for the nuclear facility. A margin of safety shall be applied for the uncertainties in the evaluation of soil and rock characteristic:.
19 Enclosure "E"
(c)
If such potential exists, a design basis shall be derived giving -
consideration to the combination of slope instability effects and the design basis seismic event.
I 100.3S Site Surface Collapse and Subsidence and Rebound (a) Geological maps and other appropriate information for the region shall be examined for the existence of natur,1 features such as caverns, karstic *ormations, and man-made features such as mines and water and oil wells. The potential for surface collapse and subsider.ce and rebound shall be evaluated.
(b) If engineering solutions appear to be a practicable solution to potential surfacc collapse or subsidence and rebound, a detailed descrip-tion of surface conditions obtained through reliable methods of investi-gation shall be developed fcr the derivation of a design basis.
I 100.36 Other Natural Phenomena Historical data conca ning pheonmena that have potential te produce adverse effects on the safety of the nuclear facility such as strong winds, severe preciptiation, snow, ice and hail shall be collected and evaluated.
If,the potential is confirmed, design basis events shall be derived using appropriate safety margins.
EVALUATION PROCEDURES FOR EXTERNAL MAN-INDUCED EVENTS r 100.40 Aircraft Crash (a) If the evaluation shows that there is a poter.tial for aircraft crash on tne site which can affect the safety of the nuclear facility then 20 Enclosure "E"
a detailed evaluation of.the probability of occurrence and of the magnitude of the consequences shall be made and the design basis aircraft crash derived.
(b) Design basis aircraft crash events shall include the effects of impac' and fire.
. 5 100.41 Chemical Explosions and Toxic Releases (a) Activities.in the region involving the handling, processing, transporting and storing of chemicals having a potential for explosion, 1
the production of explosive, gas clouds, or toxic releases shall be identi-fied'and the design basis derived.
(b) Design basis for chemical explosion events shall take into account the effects of distance on pressure forces and on the potential fur debris or missiles _ which may become airborne as a result of the explosion.
2 100.42 Ultimate Heat Sink (a) In the design of the arrangements for long term heat removal from a nuclear reactor core, site-related parameters such as the following shall be considered:
(1)' both the dry bulb and wet bulb air temperature; (2) available flow, minimum level, and duration of minimum level of safety-related sources of coolant.
(b) Suitably conservative values for these parameters shall be derived for the design basis for the ultimate heat sink.
21 Enclosure "E"
(c) Potential mane !rduced events that could cau'se a loss of function of systems required for the long term heat removal from e nuclear reactor core, such as river blockage, reservoir depletion, ship collisions, oil spills and fires, shall be identified.
If the probability and censequences of such events cannot be reduced to acceptable levels, then such avents shall be considered in the design basis for the nuclear facility.
I 100.43 Physical Security (a) The controlled area of a proposed site shall be examined for the potential of optimizing the physical security of the nuclear facility against man-induced eveats such as sabotage, riott, and intrusion.
(b)
If possible, natural protection of a proposed site for physical security shall be kept intact during the ccnstruction and operation of a nuclear facility and engineered protection shall be used to supplement such natural protection if required.
I 100.44 Other Man-Induced Events The region shall be examined for facilities that store, prccess, transport and otherwise handle toxic, corrosive, or radioactive materials which, if released, could adtersely affect the safety of the nuclear facility.
If the p'robability and consequences of such events cannot be reduced to acceptable levels, then such events shall be considered in the design' basis for the nuclear facility.
iVALUATION PROCEDURES FOR ENVIRONMENTAL ASPECTS 5 100.50 Meteorology and Atmospheric Dispersion (a) A description of~the region shall be developed including:
22 Enclosure "E"
(1) the general climate (2) summary of meteorological parameters including wind speed, wind direction, atmospheric stability, temperature, pre-cipitation, humidity and fog. As far as possible these shall shall be measured simultaneously and intercorrelated.
On the basis of these data, frequency distributions f - the above parameters shall be obtained..
(3) maps showing detailed topographic features in the vicinity of the site and general topographic features for the region.
4 (b) A program of meteorological measurements shall be carried out at the site before operation of the nuclear facility and shall include instru-
~
mentation capable of measuring and recording the major meteorological parameters at appropriate elevations and locations.
(c) For a preliminary evaluation, data from nearby meteorological stations may be used. However, if correlations with preliminary measure-ments made on the site in the various meteorological conditions are not available, the meteorological parameters so derived shall include aporo-priate factors of conservatism to account for uncertainties.
Evidence shall be provided to confirm that these data adequately represent long term conditions at the site.
(d) To evaluate the influence of unusual conditions, such cs con-trasted reliefs and thermal interferencc, additional programs shall be 23 Enclosure "E"
carried out, including in-situ diffusion experiments with tracers or simulation on scale-models when appropriate.
(e) From the data obtained on the site, or from nearby representa-tive meteorological stations for preliminary evaluations, as approporiate, a model shall be developed for the evaluation of atmospheric transport and diffusion of radioactive releases. This model shall be capable of evaluat-ing dilution factors for different values of the following:
(1) emission time periods (e.g., from minutes to weeks);
(2) source dimensions, heights, and forms (e.g., plume or cloud);
(3) atmospheric conditions (e.g., vertical stability and wind speed and direction); and (4) dry and wet deposition conditions.
The scope of the model shall include any unusual site conditions that may be present.
s 100.51 Hydrology.and Water D*spersion (a) A description of the surface hydrologic characteristics that may be present.
(1) location, size, shape, time variations and other hydrologic characteristics, including flow and velocity for rivers, curren'.s for lakes and seas, and silt load of all water bodies'both natural and aritifical; (2) major water control structures ;-d featuras; and 24 Enclosure "E"
(3) the locatien of the surface water intakes of major and minor users which could be affected by releases of radio-active materials.
(b) A program of surface hydrologic observations and studies shall be carried out including but not restricted to:
(1) determination of flows, levels, quality and dispersion characteristics of all water bodies which can potentially become contaminated or which will be used as a water supply for the nuclear facility; (2) study of the reconcentration ability of sediments and biota ir. the hydrosphere; and (3) determination of trar. sport mechanism of radionuclides through the hydrosphere and indication of exposure pathways for the significant radionuclides.
(c) The result of the above mentioned investigations shall permit an evaluation of possible impact of surface water contamination on the popu-lation. Methods to be used shall be based on historical data of hydro-logic phenomena, where they are available, and on statistical evaluations of data collected both in laboratory and field conditions.
(d) Historical data on flow rates, water levels and other parametars mentioned aoove snall be used for preliminary evaluation.
(e) A summary description of the hyarageology of the region shall be developed including:
25 Enclosure "E"
(1) description of local stratigraphy of nonsaturated zones and of water-bearing formations; (2) water table contours and their variations with increasing or decreasing hydrometeorological activity, indication of the directions of ground water movement and its velocity; (3) location of ground water sources of supply and description of their use; and (4) pathways of ground water reaching man, interaction of ground water and surface waters.
(f) Particular situations shall be identified and evaluated with respect to site suitability including high ground water levels and the presence of important points of water use in the vicinity of the proposed site.
(g) Hydrogeological investigations shall be carried out to determine that tne site characteristics and the proposed nuclear facility design permit the evaluation of the effect of release of radioactive materials at points of water use. These investigations shall include:
(1) permability and retention characterisites of soils, minerals and rocks representing the local geological structures; (2) the porosity, granulemetry, other physical and chemical properties of underground material in relation with sig-nificant radionuclides which can be released; and 26 Enclosure "E"
(3) dispersion characteristics of underground water bodies and relative velocities of the significant idionuclides which can be released.
In case of complex and difficult conditions, such as heterogenous forma-ti'n and unsteady wati' flow, laboratory studies in simulated conditions which could be verified in situ using tracer techniques may be necessary.
(h) On the basis of results of hydrogeological investigations, a model shall be developed to describe the transport mechanism of radio-nuclides through aquifers with respect to establishing the pathways through which the human population could be exposed to released radio-active materials. -
5 100.52 Porulation Distribution (a) Detailed information on existing and, if possible, projected distributions of the hute, population, including transient and resident, The in the region of the proposed nuclear facility, shall be collected.
radius within which data art collected shall be chosen on the basis of the human population distribution, taking into account special situations.
(b) The most recent census data for the region, or information obtained by extrapoitlion of the most recent census data, shall be used.
In the absence of reliable data, a special study shall be performed.
(c) The information shall include the human population by sector and by annular zones in the" region.
(d) A study shall be performed for the critical organisms of potentially affected ecological and biota systems in the neighboring area and the region of the proposed nuclear facility.
27 Enclosure "E"
5 100.53 Land and Water Uses Uses of land and water bodies in the region shall be characterized and include:
(1) lands devoted to agr.icultural uses, their acreage, princi-pal food products and yields; (2) land devrted to dairy farming (goats and cows), their acreage and yields; (3) water bodies used for com,nercial and sport fishing, including enumeration of species, their abundance and yields; (4) water bodies used for commercial and recieational purposes; l
(5) water bodies and land supporting wildlife; and (6) locations of water intakes for consumptive uses.
I 100.54 Baseline Radioactivity Baseline radioactivity of the atmosphere, hydrosphere and lithosphere for the region shall be determined prior to any construction activities at the selected site.
If possible, the prediction of future development of baseline radioactivity should be presented.
I 100.55 Baseline Ecological Conditions Baseline ecological conditions for the region shall be determined prior to any construction activities at the selected site.
If possible, prediction of future conditions should be presented for the nuclear facility in operation.
28 Enclosure "E"
BACXFITTING 1 100.60 Backfitting (a) The Commission may, in accordance with the procedure specified in this chapter, require the backfitting of a nuclear facility if it finds that such action will provide substantial, additional protection which is required for the public health and safety, quality of the environs, or the comon defense and security. As used in this section, "backfitting" of a nuclear facility means the addition, elimination or modification of structures, systems or components of the nuclear facility after the construction permit has been issued.
(b) Nothing in this section shall be deemed to relieve a holder of a construction permit or a license ~ from compliance with the rules, regula-tions, "or order of the Commission.
ENFORCEMENT 5 100.70 Violations An in,juction or other court order may be obtained prohibiting any violation of any provision of the Atomic Energy Act of 1954, as amended, or Title II of the Energy Reorganization Act of 1974, or any regtiation or order issued thereunder. A court order may be obtained for the payment of a civil penalty imposed pursuant to section 234 of the Act for violation of section 53, 57, 62, 63, 81, 82,101,103,104,107, or 109 of the Act, or section 206 of the Energy Reortanization Act of 1974, or any rule, regulation, or order issued thereunder, or for say violation for which a 29 Enclosure "E"
license may be revoked under section 186 of the Act. Any person who willfully violates any provision of the Act or any regulation or order issued thereunder may be guilty of a crime and, upon conviction, may be punished by a fine or imprisonment or both, as provided by law.
b 30 Enclosure "E"