ML17340A886

Affidavit Re Contention 2.In Applying ALARA Concept,Nrc Uses Qualitative,Programmatic Approach Rather than Imposing Quantitative Operating Limit.Prof Qualifications Encl

ML17340A886
Person / Time
Site:	Turkey Point
Issue date:	03/19/1981
From:	Nehemias J Office of Nuclear Reactor Regulation
To:
Shared Package
ML17340A884	List: ML17340A883 ML17340A885 ML17340A886 ML17340A887 ML17340A888 ML17340A889
References
ISSUANCES-SP, NUDOCS 8103240414
Download: ML17340A886 (26)

Text

UNITED STATES OF 01ERICA NUCLEAR REGULATORY COl1t1ISSION BEFORE THE ATOtlIC SAFETY AND LICENSING BOARD In the l1atter of FLORIDA POWER At(0 LIGHT COt1PANY (Turkey Point Nuclear Generating Unit Vos.

3 and 4)

Docket Nos.

50-250-SP 50-251-SP (Proposed Amendments to Facility Operating Licenses to Permit Steam Generator Repair)

AFFIDAVIT OF JOHN V.

NEHEMIAS ON CONTENTION 2 I, John V. Nehemias, being duly sworn, state as follows:

l.

I an employed by the U.S.

Nuclear Regulatory 'Commission as a

Senior Health Physicist in the Division of Systems Integration, Office of Nuclear Reactor Regulation.

2.

In its Order designating the contentions at issue in these proceedings, the Board questioned whether the cumulative occupational exposure expected during the proposed steam generator repairs is "as low as is reasonably achievable" (ALARA) consistent with the requirements of 10 CFR Part 20.

See Order Ruling on Petitions of t1ark P.

Oncavage, dated August 3, 1979 at page 49.

The staff position on this matter follows, 3.

The ALARA Conce t Section

20. 1(c) of the Commission's regulation states, in part:

[Ijn addition to complying with the requirements set forth in this

part, make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to unrestricted

areas, as low as is reasonably achievable.

The term "as low as is reasonably achievable" means as low as is reasonably achievable taking into account the state of technology, and the economics of improvements in relation to benefits to the public health and

II

safety, and other societal and socioeconomic considerations, and in relation to the utilization of atomic energy in the public interest."

The provisions of 10 CFR Part 20. 1(c) were first formally implemented for control of'ccupational radiation exposures, by issuance of the following Regulatory Guides

( RG) addressing the ALARA concept:

8.8 "Information Pelevant to Ensuring that Occupational Radiation Exposures at JJuclear Power Stations Will Be As Low As Is Reasonably Achievable."

8.10 "Operating Philosophy for J1aintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable."

8. 18 "Information Relevant to Ensuring that Occupational Radiation Exposures at <Jedical Institutions Will Be As Low As Is Reason-ably Achievabl e."

A series of similar ALARA guides is planned for other types of Li-censees.

Of these, R.G. 8.8 is directly applicable to nuclear power plants; many of its provisions relate directly to various aspects of repair of the steam generators at Turkey Point.

Florida Power and Light has committed, in its "Steam Generator Repair Peport",

to follow the pro-visions of Regulatory Guides 8.8 and 8. 10.

The former specifically requires a design review of each activity potentially involving signifi-cant levels of exposure, from the point of view of maintaining radiation doses ALARA.

Part of that ALARA review constitutes an assessment of doses likely to be caused by each such activity, taking into account the numbers of workers involved, the expected working times, and expected dose rates in each working area,'in accordance with the provisions of Regulatory Guide

8. 19, "Occupational Radiation Dose Assessment in Light-Water Reactor

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Power Plants Design. Stage Han-Rem Esti'mates."

That review inherently requires consideration of each dose-causing step in every activity involving potentially significant doses, and reasonable steps that can be tal en to reduce exposures.

In summary, the ALARA concept can be described most simply as an ongoing requirement that HRC applicants and licensees consider the radiation dose implications at eacn step of every activity potentially involving significant radiation dose rates.

In the course of planning, design, construction, operation, naintenance,

repair, replacement, decommissioning, they must think through each such step, with a view to eva'luating possible dose-reducing

actions, and implementing those that are reasonably achievable.

Regulatory Guides can list (a) examples of some of the more sig-nificant dose-causing activities, and (b) examples of possible dose-reducing

changes, but can never provide a complete catalog that would cover all possible specific situations.

The final ALARA responsibility must rest with the applicant or licensee:

to consider condi tions and situations

expected, or known to be present, in a particular licensed activity, and to take appropriate dose-reducing actions.

4.

The ALARA Process Regulatory Guides 8.8 and 8. 10,provide guidance for assuring that radiation exposures for normal operations are maintained as low as is reasonably achievable.

Since the issuance of. R.G. 8.8 in July 1973, our review of reactor applications has particularly stressed the importance of applying the ALARA concept.

Me use a qualitative,

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programmatic

approach, as distinguished from imposition of a quantitative operating limit as a criterion to judge whether radiation exposures will be as low as is reasonably achievable.

Special attention is given in R.G; 8.8, and in our licensing re-views, to:

1.

management policy and organization; 2.

personnel qualifications and training; 3.

design of facilities and equipment; 4.

radiation control

program, plans, and procedures; and 5.

availability of supporting equipment, instrumentation, and facilities.

The applicant for a construction permit, operating license, or a license amendment is expected either to follow the guidelines of R.G.

8.8, or to propose alternate approaches which would provide an equivalent degree of.personnel protection.

An assessment of the occupational radiation exposure from predictable in-plant operational activities, as well as from non-routine occurrences, is required, consistent with the guidance of R.G.

8. 19.

The purpose of tnis assessment is to assure that exposures are to be maintained ALARA by proper shielding, design, procedures, and organization.

During the ALARA review and dose assessment processes, licensees are expected to seek out and make appropriate changes in design,.plant

layout, and radiation protection practice, for the purpose of assuring that occupational radiation exposures will be,ALARA.

Each such change is directed either to lowering radiation levels, lowering the probability of situations involving high radiation levels, or reducing the time

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necessary for operations involving high radiation levels.

During the

review, pr ior to issuance of the license or amendment, and long before technical work begins, it is not possible to know precisely how often personnel may be obliged to enter such areas, what dose rates may be encountered, or how long the necessary tasks may take to complete.

In addition, it is impossible to predict what difficulties may arise in the course of specific activities that have to be per on~ed during a

particular operation.

The basic purpose of these processes is to require licensees to think carefully through each significant dose-causing activity.

NRC staff reviews radiation protection in.ormation submitted by applicants and licensees, with particular reference to design

reviews, dose assessments, and actions taken to reduce exposures.

Generally

soeaking, a careful design review, combined with a detailed dose assessment, will identify a number of possible ways to reduce exposures.

(le examine the information provided, for completeness of the design review and the dose assessment, and for the kinds and quality of dose-reducing activities considered, and of those adopted.

5.

uantitative Limits on Collective Oose Person-Rems The NRC staff has considered, at length, arguments for and against the development of annual collective person-rem limits (in addition to the present individual dose limitations) applicable to NRC licensees in general, and to NRC power reactor licensees in particular.

lie are convinced, as discussed below, that the imposition of such limits

1

would not be in the public interest, and would not be likely to result in any significant reduction in person-rem doses.

The greatest problem in the development of a collective dose limitation is in the selection of the proper dose limit.

The principal difficulty here results from the wide variability in annual collective occupational exposure experience, from plant-to-plant, and from year-to-year at any given plant.

See, for example, the summary of occupational doses experienced at pressurized water reactors frou 1975-1979 contained in the Draft Environmental Statement (NUREG-0743),

dated

December, 1980, Table 4.2.

This wide variability in dose experience from plant-to-plant, and from year-to-year at any given plant, has been a consistent aspect of reactor experience.

It stems primarily from the inherent randomness of component and equipment failure.

Highly similar plants can differ significantly in their collective radiation dose experience.

Fuel leakage

rates, formation and deposition rates for radioactive corrosion products, and maintenance needs vary widely from plant to plant, and are impossihl.e to project meaningfully for any giv n plant or for any given year.

In addition, no matter how complete and precise our data base on failure rates and dose rates were to become, it would simply provide a

probabilistic basis for projecting frequency of occurrence for particular component failures, with the associated urgent need for repair or replacement.

Because such component failures with associated high dose rate potentials occur randomly and with unpredictable frequency and dose

rates, dose experience can be expected to vary widely from plant to

plant, and from year to year at a given plant.

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If a specific operating person-rem limit were to be selected, which happened to be in the upper quartile of the annual average doses experienced at a particular plant, it would not impact that plant's operation during most years.

However, in a year during which a number of such random failures occurred at that plant, the limit could be exceeded, and the NRC might require either:*

a.

shutting down the plant, b.

operating without performing necessary maintenance (for ex-ampl,e, by isolating the equipment needing maintenance),

or c.

obtaining a waiver from the person-rem limit.

The irony of this type of sanction is that the highest person-rem doses are frequently accumulated when the reactor is already shut down.

If the Commission requires that.the reactor be shutdown, we would have the potential for a situation where work could not be done as necessary to prepare the plant for operation at a later date.

Furthermore, such a

In general, options a

and (b) would not be in the, public,interest; option (c) mi litates against a specific person-rem l,imit.

Il

limit could well serve as.

a disincentive to shutdown for necessary maintenance in order to avoid exceeding the limit in a given year.

The Staff, believes that none of the above three options for redressing a situation in which some presumed occupational exposure limit had been exceeded would be desirable considering the unpredictability of such.fai.lures, the cost-benefit judgments already made with respect to establishing justification for licensing the plant, and the risks and costs associated.

with sudden or extended cessation of power, or with delaying necessary maintenance.

6.

Cost-Benefit Considerations The Commission's first rulemaking regarding, the ALARA concept involved numerical dose control criteria for releases of radioactive materials in effluent air and water.

In this case, it was a relatively simple task for. the Commission to show that doses resulting from reactor effluent releases had been, for the most part, maintained at a level less than a few percent of the NRC annual limit of 500 millirems ( 10 CFR 520. 105).

On the basis of convincing evidence that currently available technology and current industry effluent control practices could meet these criteria, the Commission concluded that the criteria were reasonably achievable, and would be considered'LARA.

However, the Commission's regulations also state, in 10 CFR Part 50, Appendix I, Section II(D):

In addition to the provisions of paragraphs A,B and C above, the applicant shall include in the radwaste system all items of reasonably demonstrated technology that, when added to the

system sequentially and in order of diminishing cost-benefit

return, can-for a favorable cost-benefit ratio effect reductions in dose to the population.reasonably expected to be within 50 miles of the reactor.

As an interim measure and until establishment and adoption of better values (or other appropriate criteria), the values

$ 1,000 per total body man-rem and

$ 1,000 per man-thyroid-rem (or such lesser values as may be demonstrated to be suitable in a particular case) shall be used in this cost-benefit analysis.

The Comaission selected an interim value of $ 1000 per person-rem arbitrarily.

The value selected represented the highest projected from among estimates developed by a number of individuals, using various assumptions and calculations, covering a range from $ 10 to

$ 1000 per person-rem.

The cost considerations involved in these estinates were limited to the costs associated with the assumed health effects, e.g.,

medical attention or working time lost.

In effect, this provision requires additional efforts to reduce concentrations of radioactive materials in.effluent air and water, but only to the extent that such further reductions could be achieved at a cost less than

$ 1000 per person-rem saved.

From a cost-benefit point of view, control of occupational radiation doses differs significantly from control of public doses, speci icially in its inherent variability from plant-to-plant, and from year-to-year at any g'iven plant.

As a result, the data provide no basis for selecting generally applicable numerical collective dose values analogous to those in Appendix I.

In Regulatory.

Guide 8.8, reactor applicants and licensees are encouraged to make use of cost-benefit analyses in selecting among alternative ways to go about reducing doses, and in deciding whether or not to take a particular dose-reducing action.

The Commission has not as

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yet, however, recommended any specific mode of calculating costs and

benefits, nor of quantifying the values of parameters involved.

The International Commission on Radiological Protection has developed a cost-benefit methodology in Publication 22, whicn provides some guidance as to such analyses.

British and Canadian authori ties are proceeding with development of similar processes.

To date,

however, none of these efforts have yielded fully quantitative, reliable results.

Advantages of proceed'ing directly with the development of such a

consensus relate to the positive value of having available a functioning, agreed-upon cost-benefit procedure.

The process could thereafter be applied uniformly and routinely to each significant dose-reducing activity under consideration.

A straight-forward quantitative calculation could then be used to determine, case-by-case, whether or not a particular activity would be ALARA, and, therefore, whether or not it should be undertaken.

Difficulties in development of a consensus on such a course of a"tion are

( 1) the serious difficulties and complexities of seeking agreement on a process to determine parameters to be included, and appropriate quantitative values to be selected; (2) the substantial cost in resources and in personnel (including NRC staff as well as industry personnel) that would be involved; and (3) the possibility that strict

Ct application of such a quantitative cost-benefit equation could very well result in a determination that many widely-accepted current radiation protection practices would'ot be found to be justified.

own V. Nehemias Subscribed and sworn to before me thisjJ "day of Narch, 1981.

<rw Notary Publ i;..

Hy Commission expires:

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John V. Nehemias PROFESSIONAL QUALIFICATIONS Radiological Assessment Branch Division of Site Safety and Environmental Analysis I am a Senior Health Physicist in the Radiological Assessment

Branch, Division of Site Safety and Environmental Analysis, Office of Nuclear Reactor Regulation.

Ny formal education consists of study in Physics at Rensselaer Polytechnic Institute where I received a B.S. in 1948 and at Columbia University where I received an A.N. in 1949.

I received a Ph.D. in Environmental Health (Radiological) from the University of Michigan in 1960.

Before joining AEC/NRC, I served three years at Brooknaven National Laboratory as a health physicist, six years at the University of Michigan as health physicist: and adsistant director of a radiation ef ects laboratory, and three years as Director of Radiological Health Surveys for the National Sanitation Foundation.

In the latter position, I

designed, organized, and directed the environmental survey for the Enrico Fermi nuclear plant.

I joined the AEC in September

1960, as a health physicist in the Office of Health and Safety.

f)y principal duties there related to development of radiation protection standards.

Mith the two exceptions noted

below, I have continued with AEC (and NRC) since that time.

l<y principal responsibility was in the development of Stanoards until September 1974; during most of those years I served as a branch chief-through several name changes and reorganizations-most recently as Chief, Occupational Health Standards

Branch,

>~larch 1972 to September 1974.

Since September

1974, I have served as Senior health physicist in the Radiological Assessment Branch.

Ny principal function is the review of power reactor applications, both at the construe.ion permi't and operating license

stage, to determine the adequacy of proposed occupat onal radiation protection programs and the related efforts proposed to assure that occupational radiation exposures wi 11 be maintained as low as is reasonably achievable.

it From June 1963 to SeptemGer

1965, I took a leave of absence from AEC and served as principal member of the Occupational Safety and 'Health Division of the International Labor Of ice in Geneva, Swi zerland.

Hy work was principally in the development of international standards.

In December

'1971, I was trans,erred to the Criteria and Standards

Division, EPA, serving as Chief, Criteria and Standards Brarch, until my return to AEC in Farch 1972.

I have published about 40 technical articles in professional journals and other publications in the general areas of 'low-level counting, environmental monitoring, radiation effects on biological systems, and c"n rol OT occupation 1 radiation exposul e.

I have been a Cer.ified Healtn Physicist since

1960, and am a Charter member of the Health Physics Society and of the Baltimore-Washington Chapter.

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