ML20138A273

Forwards Answers to 850808 Questions Re Proposed Rev to 10CFR20, Stds for Protection Against Radiation

ML20138A273
Person / Time
Issue date:	10/18/1985
From:	Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	Asselstine NRC COMMISSION (OCM)
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ML20138A277	List: ML20138A273 ML20138A287
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NUDOCS 8512110639
Download: ML20138A273 (8)

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. . . ,,9 cc1CDG certi m OCT 181985

,-(P MEMORANDUM FOR: Comissioner Asselstine FROM: William J. Dircks Executive Director for Operations

SUBJECT:

PROPOSED REVISION OF 10 CFR PART 20, " STANDARDS FOR PROTECTION AGAINST RADIATION" (SECY-85-147)

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In your memorandum to me dated August 8,1985, you asked four questions about the proposed revision of Part 20. Our answers to these questions are

. provided in the enclosure.

(Sipec William J.Sireks William J. Dircks Executive Director for Operations

Enclosure:

As stated cc: Chairman Palladino Comissioner Roberts Commissioner Bernthal Commissioner Zech OGC OPE

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! 1.Q. The proposed revision to Part 20 would establish a dose limit to a j == amer of the general pubite of 0.5 rem per year. Page Z1 of [

Enclosure 1 of su,Y-55-147 indicates that that dose,11mit translates to l 1 a calculated annual risk for an individual of 5x10 per year, which is i l

deemed on page 18 to be an acceptable level of risk. Un the other hand.  !

the objective of the comission's provtsional safety goal is "to establish L coals which limit to an acceptable level the radiological risk which

! might be imposed on the public as a result of nuclear power plant ,

operation. Including normal operation, expected transients and design l basis accidents. The quantitative goal for latent cancer risk of an  ;

il individual is about 2x10 per year (see page 31 of NUREG-0850  ;

! Revis1on 1J. What 1s the rationale for two " acceptable" levels of i risk that differ by a factor of 407 Why should not Part 20 be revised l to conform to the provisional safety goai7 l

- l 1 A. The dose limits in 10 CFR Part 20 and the Connission's Safety Goal were i

! developed for different purposes and are not directly comparable. The 500 arem/yr limit for members of the public includes dose from all sources of l exposures except background radiation and medical treatment. It is not  ;

j limited to the exposure resulting from a single source as is the case for the Safety Goal value. Furthemore, consistent with the recomendations or guidance from ICRP, NCRP, and FRC and required in the pro ssed 10 CFR l l

20 Part, radiation dose must be maintained as low as reasonally achievable l

- (ALARA). ALARA is the operative mechanism by which acceptable radiation i J programs are demonstrated. 10 CFR Part 50, Appendix I and 40 CFR Part 190 l

! are specific regulatory ALARA requirements which take into account what can be achieved through the application of good technology and procedures.

Appendix I was developed by the Comission, after rulemaking hearings (RM l 50-2) to supplement the dose limits of Part 20 by adding design l

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objectives to assure that gaseous and liquid effluent releases to the i environment from light water power reactors not only were within Part

20 limits but were also consistent with ICRP, NRCP and FRC guidance that radiation exposure should be kept "as low as practicable".

The staff response to Appendix ! was to include Radiological Technical .

Specifications (RETS) as a part of each plant license. The RETS require j that all licensed power plants operate within the guidelines specified in Appendix 1. For routine releases of gaseous effluents, releases are limited during any calendar year to those which could result in an air j dose at the site boundary of 10 m1111 rad genna or 20 millirad beta *.

Exceeding this specification does not require imediate action, but does i

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• These 11mits are cast in tems of air dose measured in m1111 rad; Appendix I l also provides that the Comission meTspecTTy a lower release 1f releases

! resulting in air doses at the 10 millirad game,20 millired beta level are likely to result in an estimated annual dose to an offsite individual in excess of 5 millirem whole body.

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1 require a report to the Conunission identifying corrective actions taken or to be taken. Quarterly specifications are also imposed to assure that i

corrective actions are taken if the releases in any quarter reach one-half l

the annual air dose design objective. These requirements assure that j long tern routine gaseous releases are kept to fractions of the Part j 20 limits for power reactors. ,

.I l In addition, to protect against off normal releases that might be i

associated with transients or malfunctions, the Tech Specs also contain a i requirement limiting "at all times" gaseous releases to a rate of 500 l

milliren/yr (a rate which, if continued for a year, would result in a dose  !

at the site boundary of 500 millirem; e.g. an hourly rate of approximately j 0.060 millirenVhr). A deviation from this limit requires inmediate action '

j to reduce the release rate to within 500 millimrem/yr. Thus, even off i

- nonnal releases would be controlled under the Tech Specs to small i fractions of the Part 20 limits.*

l This combination of limits thus provides for protection for a wide  ;

i spectrum of release conditions from routine anticipated releases to

! unanticipated releases resulting from transients a; d malfunctions,

assuring that insnediate action is taken in the event of transients or l' malfunctions to limit the rate of effluent release to a rate of 500 l millires/yr while still assuring that the annual quantity of material  !

released is within the Appendix I design objectives of 10 millirad gamma

< and 20 millirad beta. This conbination of limits also assures that

! effluents from power reactors satisfy the EPA standards in 40 CFR190. i i Based on a review of operating experience over the last decade, all power i plants have operated with the levels specified in Appendix I except one,

Rancho Seco in California. Although the Rancho Seco experience is 1

presently under review, it appears that an administrative error may have -

caused the plant to exceed Appendix I from 1980 to 1984. We conducted a l detailed survey of potentially affected offsite areas and are currently establishing possible exposure levels. Rancho Seco is now operating  !

within Appendix I limitations.

, - Limitations on routine releases are not directly comparable to safety goal

! objectives. Routine release limits are based upon doses and dose projections for points at the site boundary (or for any individual for the

! 119 id pathway). The safety goal objective for latent cancer risk, to i Ql3 the question is directed, relates to the risk to the nearby l

populatic.. Thus, one is attempting to compare point (or individual) risk to population risk. Nevertheless, even if a power plant operates at the i dose design objective values, an individual (whose exposure were limited i

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• E.g. assuming a malfunction or error resulted in a release rate, for a ganna ,

l emitting isotope,10 times the limit rate and took as long as four hours to l i correct, the release for that period would still amount to only about 2.5 mrem l at the site boundary if meteorological conditions were average; this would, l moreover, use up half of the quarterly limit so that if additional routine l releases in the quarter amount to over 2.5 millirem, a report identifying l appropriate corrective action would be required.

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I to 5 millirem per year, in accordance with A an increased risk of cancer of about 8x10"7 . pp.I.Sec.

This risk is aboutIIB 40%2) would of incur the safety goal.*

!' While the health and safety considerations associated with routine reactor effluent releases and the Commission's safety goal objectives are not directly comparable, the actual Commission program for control of routine j releases assures that such releases are consistent with Appendix I, consistent with 40 CFR 190, and are well within the Commission's safety j goal objectives.

l The purpose of the Appendix I rulemaking was to establish objectives to implement the general principle of radiation protection that not only 1 should exposures be kept within applicable limits but also that exposures j

to the public should be kept "as low as practicable" ("as low as reasonably achievable"). Proposed Part 20 makes much clearer the obligation of all licensees to develop and to implement an ALARA program.

" The commission's safety goals do not include a quantitative design objective for the latent cancer fatality risk to any specific individual; the cancer fatality design objective applies to a population (defined as those persons within 50 miles) and states that this risk should not be

increased by more than 0.1 percent compared to the sum of cancer fatality

! risks resulting from all other causes. The risk fatality in the U.S. is on average about 1.9x10~3 estimate for a cancer which means that the design objective (rgunded would up)per be equivalentMrsonto per year, an increased latent cancer risk of 2x10- per person per year, on the average, for persons within 50 miles of the nuclear plant.

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l 2.Q. If one accepts the proposition on page 47 of Enclosure 1 that the dose

' limit to the general public of 0.025 rem per year established in 40 CFR 190 was based on ALARA considerations, what is the rationale for proposing in Part 20 a dose limit that is 20 times higher than that which is reasonably achievable 7 A. The EPA 40 CFR Part 190 dose value is an ALARA perfonnance standard, whereas the value of 0.5 rem per year in the present and proposed 10 CFR Part 20 is not an ALARA perfonnance standard. The Part 20 dose values are limits, based largely on health-risk estimates, intended to provide universally acceptable levels of health protection consistent with national and international recomendations (National Council on Radiation Protection and Measurements, International Comission on Radiological Protection, and the fonner rederal Radiation Council). These limits represent an upper bound to allowable dose levels under nonnal operation l

conditions and apply to all classes of NRC-licensed activities.

The EPA 40 CFR Part 190 dose value is an ALARA perfonnance standard which is technology-based for uranium fuel cycle facilities. As such, it is not

' applicable to all classes of NRC licensees. Furthennore, the ALARA standards, like EPA's 40 CFR Part 190 and NRC's 10 CFR Part 50, Appendix I, represent perfonnance levels achievable below the regulatory health-based limits with the state-of-the-art equipment and systems and

optinni operation of this equipment and systems. 10 CFR Part 50, j Appendix I provides operating flexibility by pennitting operation above the stated levels for limited periods of time while corrective action is taken. Unlike the health-based limits of Psrt 20, costs are a primary l

consideration in the development and implementation of ALARA standards.

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Annual dose values well below the regulatory health-based limits are generally achievable at reasonable costs for normal operations of uranium fuel cycle facilities. In developing its 40 CFR Part 190 i standard, EPA drew upon the same studies and operating experiences used by NRC to develop Appendix I.

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3.Q. SECY-85-147A indicates that the British have adopted a "de minimis" standard of collective doses less tnan 100 man-rem comprised of individual doses less than 0.5 mrem per year. What is the rationale for not proposing in Part 20 a societal "de minimis" standardT A. Our understanding of the intent of this question is based or, i-clarification obtained from John Austin by telephone on August 21, 1985.

Noting that a recent British NRPB recommendation would allow the use of a cutoff in collective-dose calculations only where collectiye doses of 100 personrems or less are involved. Commissioner Asselstine would like to i know the staff's rationale for omitting a collective-dose, criterion of 4

this nature from its recommendations. Our rationale is presented below.

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- The proposed 1 millirem per year cutoff level, which represents less than I percent of the average natural ba k associatedwithariskofabout10"9greyndradiationdoserate,isper ytar. -The ave course, be much less. The staff evaluated this small risk from the individual and societal viewpoints. Risks associated with I' millirem per year or less should be of no concern to the exposed individual and normally would not affect any decisions that he or she may need to make.

As long as no individual receives more than 1 millirem, the risk should be 1 of no concern to any individual and the collective dose would inherently 4

be limited to levels of little concern to society. On this basis, the j staff sees no need for imposing an additional limit on collective doce.

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! 4.Q. Can the staff certify that this Part 20 rulemaking represents a substantial increase in the overall protection of the public health and safety or the comon defense and security? If not, on what basis does 5 the s;aff believe this rulemaking would be in compliance with the backfit rule 7 The staff cannot certify that the proposed Part 20 rulemaking constitutes the j "backfitting of a facility" from which there is derived "a substantial increase l

in the overall protection of the public health and safety or the common defense and security" as those tems appear in the new backfit rule (10 CFR 50.109(a)(3).

While the matter is far from crystal clear, the Part 20 rulemaking with its

,s generic application to thousands of licensees and not just the relative handful ,

e of reactor licensees does not fit well into what the Comission seems to have

had in mind when it was dealing with the "backfitting of a facility." The j .. statement of considerations which accom>anied final promulgation of the backfit i rule (50 FR 38097-38111) is replete wit) reference to " future management of j backfittGg for power reactors," both plant-specific and generic changes as j applied to one or more classes of power reactors," and " modifications to a j nuclear power plant." Moreover, the regulation itself defines the tem i "backfitting" in terms of

l - modification of, or addition to, systems, structures, components, or 1 design "of a facility "

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! - modification of, or addition to, the design approval or manufacturing F1 license "for a facility," and f - modification of, or addition to, the procedures or organization required

! 'to design, construct or operate a facility."

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! It can, of course, be argued that the proposed changes to Part 20 represent a i " modification" of the " procedures" necessary to " operate" a facility. However, it seems to stretch the plain meaning of the words inasmuch as a facility's

operating procedures, while developed within an overall environment of which j general radiation protection standards represent an important part, do not i

establish radiation exposure standards themselves. Operating procedures do not

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establish permissible dose levels for people or permissible radiation concentrations. Rather, operating procedures take the types of standards set

! forth in Part 20 as given and go on from there.

j The question under discussion here would be resolved, of course, if it were i clear one way or the other whether the Comission intended that the application of the backfit rule to "rulemaking" meant all rulemaking of any sort j wratscover, on the one hand, or merely rulemaking which has an effect on power i . reactors, generally, similar to the effect a backfitting order would have on a

single reactor, on the other hand. I am persuaded by the " legislative history" l of the rule and the very words of the rule itself that the better view is the

latter interpretation. Therefore, in response to the second question,

! coirpliance with the backfit rule in this instance is not necessary.

A. Notwithstanding the foregoing, the staff believes that the proposed revision of Part 20 will provide a sound scientific basis for the NRC i Standards for Protection Against Radiation and will adjust dose limits for i

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a workers and the public to be consistent with contemporary radiation protection philosophy. It also incorporates up-ta-date radiobiological ,

information which will cause the dose estimate to more closely reflect the i

actual dose from a given intake. The revision will also require licensees to keep exposures "ALARA", rather than er.horting them to do so as in the 1, present regulation. -

For some licensees (fuel fabricators), the revision will require >

i control of worker exposure than the present Part 20; for others (greater nuclear l l

power plants) there will be very little change. For many workers and j members of the public, there will be greater confidence in the adequacy l

and accuracy of the level of protection which is provided. The EPA has j circulated, to affected Federal agencies, proposed Federal guidance for occupational exposures. This guidance would be implemented by the proposed revision of Part 20. It has been the policy of the NRC (and the AEC before that) to implement such guidance when signed by the President.

There have been no cost-benefit reservations attached to this l implementation in the past and it is not clear whether the Comission i intends for this to be the case in the future. If there are to be such.

reservations based on the "backfit" rule, a major policy issue would appear to be involved.

The syst'am of dose limitations and other changes in the proposed revision of Part 20, such as modifications in the pennissible amount of intake for

! some radionuclides are based on health protection considerations and

, confonnance with Federal radiation guidance and the reconmendations of the l ICRP and the NCRP. Consequently, cost-benefit justifications

(optimization analyses) are not germane.

The revision specifies a maximum annual dose (500 mrem /yr) for the most

! exposed individual and includes a more restrictive dose limit for .

chronically exposed individuals (100 mrem /yr " reference" level). Again, i these changes are based on health effects considerations, not ALARA, and

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should not be subject to the cost-benefit considerations which are j sometimes used to justify "ALARA" requirements.

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