ML20028F951
| ML20028F951 | |
| Person / Time | |
|---|---|
| Issue date: | 10/03/1982 |
| From: | Minogue R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| Shared Package | |
| ML20027A699 | List: |
| References | |
| FOIA-82-530 NUDOCS 8302070211 | |
| Download: ML20028F951 (20) | |
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[C NRC ACTIVITIES IN THE REGULATION OF. RADIATION by Robert B. Minogue, Director.
Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission
&2 Washington, DC 20555 gn AIF Conference on Radiation Issues for the Nuclear Industry Royal Sonesta' Hotel New Orleans, Louisiana 3-6 October 1982 8302070211 821221 PDR FOIA SHOLLYS2-530 PDR
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NRC ACTIVITIES IN THE REGULATION'0F RADIATION
- by Robert B. Minogue, Director Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington, DC 20555 The nuclear industry and the NRC both are concerned with providing the necessary pro "ction of public health and safety for the complex plants-that you design, construct and operate, and that we regulate. - Because we-have this shared responsibility,' and want our activities < to be' effective',,
it is essential that we have meaningful dialogue on issues-that affect this protection.
To be a meaningful dialogue the -information exchanges must reflect the best scien e and engineering, must be candid, and most importantly must be conducted in an atmosphere of mutual respect, one which recognizes-our respective responsibilities to the public interest. We must. listen open-mindedly to each other and to others who have a voice in the matter.
In this spirit of cooperation, the decisions we reach should be technically sound, reasonably priced. socially acceptable and not counter-productive to the protection we both seek to provide.
o It is in this spirit, I believe, and to provide an opportunity for such dialogue, that the Atomic Industrial Forum convened this conference on radiation issues.
I very much welcome the opportunity to speak briefly on several issues that I believe are important and particularly timely.
- Presented at the AIF Conference on Radiation Issues for the Nuclear Industry, 3-6 October 1982, Royal Sonesta Hotel, New Orleans, Louisiana.
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,. The first issue is that of the source term-for radioactive material which might be released under reactor accident conditions. Knowing the source term under various accident scenarios is basic to-the fomulation of well-founded engineering practices and sound operations and to our ability to regulate based on realistic assessments rather than relying on arbitrarily chosen " limiting" cases which are somehow assumed to characterize a wide range of accidents.
For the second issue I will provide a 'few remarks on our current efforts to revise our basic radiation protection standards, Part 20. Much has changed since Part 20 was originally published in the late 1950's that warrants our examination of this basic regulation.
This revision is an attempt to better reflect the present philosophy of providing radiation protection in the work place and for the gen'dral public, and the developments and a'dvances in related sciences useful to implement that philosophy. You will hear more on it from Dr. Mills later in the conference.
The third issue is that of defining a level of health risk which is so small that neither the regulating agency nor the industry should expend resources to control activities producing health risks at, or below, this level.
We are incorporating this concept which is sometimes described as "de minimis,"
into the draft revision of Part 20 and it is important that we both understand its meaning and its implicetions.
I will provide some general coments and Mr. Cunningham will describe this concept further.
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. I note also that you will be hearing from Dr. Congel on the NRC's -
thinking on the ALARA principle; this is a basic precept in our radiation protection program.
Source Term Safety reviews are highly dependent on accident source tenn assumptions.
The Comission is comitted to develop and adopt more realistic source terms.
A major effort is being put forth in the U.S. and around the world to develop needed data and to establish best estimate source tenns for projected reactor accident scenarios.
Through the sixties and early seventies the accident assessment reviews began with a series of assumptions about fission product releases from the fuel and primary coolant system in -the event of-an accident' involving fuel damage. These assumptions were intended to bound the possible range of offsite accident consequences within the spectrum of credible events. These assumptions were described in AEC Technical Information Document (TID-14844). Among these assumptions were those concerning the release of fission products to containment for the maximum credible accident; i.e., 100% of the noble gases, 50% of the halogens and 1% of the solids in the fission product inventory.
This was-the assumed character of the release to containment following a major breach of the primary coolant system and partial melting of the core.
In the early assessments fif ty percent of the iodines released to the containment building
- were assumed to remain available for leakage to the atmosphere.
These assumptions about releases to containment and depletion within containment were set forth originally as part of a set of assumptions to be used in a rigidly formal and
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. conservative manner to determine the~ minimum size of the exclusion -zone (EZ) and low population zone.(LPZ) boundaries. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (EZ) a'nd accident duration (LPZ) doses calculated using these assumptions were then compared to the reference doses of 25 rem to the whole body and 300 rem to the thyroid given in 10 CFR Part 100. Other assumptions in the TID-14844 calculation included a containment leak _ rate of 0.1% per day and specified weather conditions for the duration of the-release.
(Subsequent ~ license actions refined the calculation of the reference doses with more sophisticated treatment of meteorological conditions and design parameters but the fractional.
releases of fission products remained unchanged.)
These values were introduced'only as part of a hypothetical case.to demonstrate a procedure for specifying<the exclusion zone and low population E1 zone boundaries,,using the guideline doses of Part 100, that would be consistent with licensing experience up until then. The NRC staff and the industry, however, not only rapidly adopted them as a connon basis for structuring the siting review but also began to apply them to the review of the effectiveness of engineered safety features incorporated as a tradeoff for site features or to mitigate accident consequences. The effect was to extend these assumptions to plant design. Thus the convenience and stability afforded by having a standard source term perpetuated the use of the TID-14844 release assumptions in these areas for which they were not intended, and to which they were not fully applicable, and in some cases, not applicable at all.
i In the early 70's a source tenn reevaluation was undertaken as part of the Reactor Safety Study (RSS). This reevaluation took into account work done since 1962 at ORNL, PNL and Battelle Columbus on different areas of fission product chemistry and transport.
In the Reactor Safety Study an I
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, attempt was made to develop more realistic' source terms for well defined accident sequences, but the work on developing a phenomenological data base had not been crr.pleted.
Therefore, certain assumptions such as the chemical form of iodine were based more on historical tradition (TID-14844) than on revised analysis. Among the many controversies surrounding the RSS were observations that the source tenns were unrealistically' conservative and needed reevaluation.
After TM1 these source term criticisms became more frequent since the release at TMI was not characterized by the earlier assumptions, particularly the iodine release.*
As part of the TMI Action Plan, efforts were begun to develop a more systematic approach to accident evaluation, building on the techniques developed in the RSS and improving the data base on fuel damage-and the models of-fission product behavior in accident environments.
Increasing controversy concerning the conservatisms in the accident source' terms contributed to the initiation of a significant effort to understand the mechanisms of behavior of fission products following release from fuel that would affect the possible ultimate atmospheric release of fission products and the subsequent transport of the fission products.
Research Our esearch program now is looking both experimentally and theoretically at the chemistry and physics involved in all phases of severe accidents, that is, accidents in which there is fuel damage such that fission products escape from the fuel. We are looking at component failure sequences and system
- Although TNI core damage was not extensive enough (in terms of fuel temperature) to plausibly be related to a core melt accident.
. transients for a wide range of accidents.
The source term work is
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focusing on the mechanisms of transport or depletion of fission products as they pass from the fuel to the atmosphere under the possible conditions present in the primary coolant system and in the containment building.
Ideally, one could begin at the fuel pellets and follow the degradation of fuel to obtain a time and' circumstance dependent description of fission product release from the fuel under accident conditions. Then the transport and depletion (deposition) of these nuclides could be traced through the primary coolant system to the point of breach to see which nuclides escape from the primary system and then through containment and the accident mitigation systems in containment, and to determine in what quantities they escape.
Accident conditions Mn' containment can also be modelled to predict the conditions ps affecting-fission product transport and leading up to or causing containment failure.
The time before release from containment and the conditions in containment at the time of release are very important for characterizing the retention of fission products or the release of radionuclides to the environment.
When does containment fail? How does it fail? What and how much gets out?
Where are the radionuclides that didn't get out? The NRC research program, together with efforts of DOE, EPRI, IDCOR and inurnational research programs, is attempting to throw light on these and other related questions.
The NRC research program is concentrated at ORNL, Battelle Columbus, Sandia, and INEL.
First, continuation of fission product release experiments at ORNL will measure fission product release rates from short segments of discharged
.. commercial PWR/BWR fuel elements.
The experiments are being conducted out-of-pile in an induction furnace within a hot cell at isothermal temperatures U
U between 1300 C and 2600 C.
A second program at ORNL is examining samples of trace irradiated fission product simulators. The objectives are to determine fission prod'act release rates, rate of fuel and structural aerosol release, physical and chemical characteristics of released species, behavior of the mixed aerosols in containment and perhaps the primary system, and effects of hydrogen ignition on fission product and aerosol transport and physical and chemical characteristics. A third program at ORNL is reviewing the chemical behavior of major iodine species in aqueous solutions. The goals are to determine the reaction kinetics up to 150 C, measure partition coefficients and establish the existence of hypoiodous acid. The tests do not involve radioactive materials.
The last major effort at ORNL is the aerosol transport test program in steam in the Nuclear Safety Pilot Plant (NSPP). The objective of the program is to provide experimental data on the behavior of aerosols released into containment under assumed LWR accident conditions. Components of the LWR core melt aerosol to be considered are those species from fuel, cladding, structural materials, concrete and coolant.
Effects of moisture, in-vessel sprays and fogs, auxiliary cleanup systems, and containment equipment on aerosol behavior and removal will be studied.
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At Sandia National Laboratories fissioniproduct va]or phase chemistry is being studied to identify the chemistry and interactions which'may affect-the transport of fission products. from the fuel into the reactor containment.
-The chemistry and. transport of fission products in typical steam =and hydrogen environments will be compared with thermodynamic calculations.
Medium-scale fission prod,uct and aerosol release, transport, and deposition -
tests will be ' conducted in the Power Burst Facility by'EG&G Idaho, Inc. as part-of the NRC Severe Fuel Damage Program.
Five bundic experiments are' presently planned for Phase I in which fuel rods will be heated in; steam _to peak temperatures up to 2400 K.
Higher temperatures, potentially up to the: fuel melting point, will be achieved in a possible Phase II' program.
The major emphasis in the phasehlf pests would.be to measure fission productLtransport under prototypic conditions for severe accident-sequences. The system will include measurements 5f the relative timing of-volatile fission product and aerosol release, time-dependent measurement of aerosol release in the primary system, specific measurement of selected fission product chemical forms, discrete characterization of aerosol release and integral fission product 1and.
aerosol retention.
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Together with EPRI, NRC is also contributing to the' multi-national program at the Marviken Facility in Sweden. This activity,. currently in the planning stages, will be used to conduct large scale fission product / aerosol transport experiments. A major emphasis will be fission product attenuation within the primary system. Aerosols will be generated by vaporizing non-radioactive. fission product simulants to represent the aerosol source expected during core-melt accidents.
The primary objective is to create a large-scale data base on the behavior of vapors i
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.. and aerosols within typical LWR primary systems for the verification of the transport models for risk-dominant eccident scenarios. -The secondary objective is to provide a large scale demonstration of the behavior of aerosols in primary systems.
Tests are being considered with simulated fission products (fission) and simulated fuel-structural material (corium).
Model development programs are underway at BCL, Sandia, and ANL to develop models of fission product release from the fuel during in-vessel fuel heat-up and melting and during ex-vessel core debris interactions with concrete in the lower reactor cavity.
Development of the TRAPMELT code, which models fission product and aerosol attenuation within reactor coolant system components and piping is -
continuing at BCL.
A number of containment fission product and aerosol models are under development. These include the fast running MATADOR code _at BCL and the more mechanistic extended TRAPMELT code (BCL) and the CONTAIN code (Sandia).
i DOE and EPRI are funding the Fuel Source Term Chemical Form Identification 1
program at HEDL and ANL.
The objective of this program is to reduce uncertainties in the LWR fission product source tenn by unequivocal identification of fission product compounds and quantification of vapor pressures. Commercially irradiated fu 1 discharged several years ago will be heated in a furnace up to approximately l
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. 2800 C, and the fission product chemical species and fonns will be identified using various techniques. These data, combined with the NRC program data will be very useful in the development and assessment of models to describe fission product transport and deposition within the primary system.
IDCOR is engaged in several analytical tasks regarding fission product release, transport, and inherent retention. The data base and techniques will be assessed for predicting quantity and chemical forms of fission product release from fuel for various dominant sequences, considering effects such as burnup, heating rate, melting rate, U0 dissolution by zircaloy, and 2
chemical states.
Data and models for predicting fission product transport behavior within primary and containment systems will also be assessed.
For all aspects ofLthe work, the state-of-art will.be assessed and needed data n:2 a will be identjfi,ed, but no experiments will be performed as'part of this.
program.
EPRI is conducting an analytically based program to develop a set of experimentally validated computer programs which more realistically simulate the release of radionuclides from degraded cores, transport through the primary coolant system and containment, and release to the atmosphere. To accomplish its objectives, EPRI is conducting research on aerosol transport in the primary coolant system, water scrubbing of fission products, aerosol transport in containment, and the duration of containment integrity. Their ma.ier effort is directed toward determining fission product and aerosol reter. tion in the primary coolant system.
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. Particularly important.is the program planned by the French on the Phebus reactor, in which NRC intends to participate. The European Community.
also plans a longer term project at the Joint Research Center at Ispra in which the NRC is also a participant.
Two particularly important areas of fission product research are being conducted in Germany. 'Out-of-pile heating of irradiated fuel, to determine I and Cs release, is being performed on a small scale. ~ Out-of-pile experiments are being performed in the SASCHA Facility to measure the release behavior of fission products from simulated fuel rods at temperatures between 1500 and 2800 K.
Thus a significant world-wide effort is underway to characterize fission product. release from fuel both before and after melt conditions have occurred.
Gaseous, volatile and aerosol release components are all being studied so that their chemical and physical behavior in the primary coolant system and in containment and in the presence of steam, hydrogen and oxygen can be predicted.
Extensive modelling efforts are underway as well as experimental efforts to provide the data to validate the models.
Experiments include out-of-pile and in-pile tests and both actual and simulated reactor core materials. The program is extensive but not exhaustive.
The coupling between the experimental and theoretical ' efforts is critical if the program is to yield reliable tools for future engineering evaluations.
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, Application Upon completion of this work we will have a much better understanding of what might happen in the course of a real accident and the identity and quantities of radionuclides that may be released. We will also have a. series of simulation models that have been calibrated with experimental results and are capable of' extrapolation,to accident sequences that have not been analyzed
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in detail.
How will these tools allow us to do a better job of regulating the nuclear industry? A number of areas suggest themselves as immediate beneficiaries of the source term research.
These areas are equipment qualification, probabilistic risk assessment, and definition of siting and energency planning a>.
requirements. As soon A$ kderesults of the near term source term work are D!'
available the technicaljanalyses used to support the energency planning regulations and their implementation will be re-examined to determine whether any changes should be made in those requirements.
The development of siting criteria will be continued with 'a reexamination of the sensitivity studies to gain insight into siting parameters important to safety. Proposed regulations will be submitted to the Commission with full knowledge of the nature of any significant changes in the accident source terms.
Equipment qualification requirements will be reviewed with an eye towards the environment within which safety systems may have to operate in the event of an accident.
The range of physical conditions in the primary coolant system f
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- parameters with which a system will have to deal to accomplish.its1 function (e.g., pumping sump water with highly abrasive suspended particulates)' or.
just to keep running (e.g.,. shorting of. electrical systems, radiation damage to components in high flux areas). The equipment qualification program will thus have a much finner basis for establishing performance requirements.
I A particularly important area of application will be to risk models-and best estimate codes for risk-assessments. The present state of risk; i
assessment techniques suffers both from limitations in the' methodology and incompleteness of the phenomenological-data base. A better understanding of the phenomenology of accidents ' involving fuel damage will permit' realistic treatment and should much improve the usefulness of risk' assessnent. The revised source term data base'will be particularly useful to us in our current'-
evaluations of the effectiveness and need'for additional plant features. to reduce the risk of severe accidents. We realize that evaluations using the older, more pessimistic: source terms can significantly overstate the.
effectiveness of added features in reducing risk.
10 CFR Part 20 As noted earlier, Part 20 has been in our regulations since about 1960.
It was originally based on the recommendations of the International Commission on Radiological Protection (ICRP) Publication 2 (1959) and the-
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4 Federal Guides approved by President Eisenhower in 1960.- Many changes have i
been made in Part 20 since -1960 to keep it current, but several notable 1
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~ deficiences remain, such as the maximum permissible concentrations of radionuclides ~ in air and water being basedion outdated,and erroneous.
. biological information.
This results in calculated organ doses which are' inconsistent with the latest scientific information.
An obvious shortcoming has been that the protection limits provided in Part 20 had no clear rationale behind them.that -explicitly. expressed the protection provided; in essence-the limits can be construed to have been' selected on the basis that no harm was expected. With the publication
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of'ICRP 26, a system of protection' that included annual limits based on estimates j
of risk to health was formulated and recomended.
H I believbderiving protection standards on the basis of ' estimate's of health risks h'is' great merit.
Even though we don't profess to have all c
possible knowledge on the biological effects of ionizing radiation, we do have, as the result of decades:of research on the subject, a vast amount of knowledge that allows us to make reasonable estimates without expectation
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of significantly underestimating any detriment to health. Thus.we can regulate more effectively and provide the worker and the public with a perspective on health protection, and with a better understanding of-why we set limits where we do.
In addition, health risk-based standards have the advantage of internal consistency and comparability with other sources of risk'and are inherently consistent with the NRC's move to make' more use wa 4
' of risk assessments-in decision-making.. The use of a health risk es'timate is -
an integral.part of NRC's probabilistic risk analysis effort and considerations 4
of safety goals. Why not make it a similar.part of our basic protection standards?
A drafting group has developed a number of draft revisions; based,- in-part, on the reconrnendations of the ICRP and, in'part, on. specific needs identified to make the ICRP system compatible with U.S. -interests. Many-of you I expect have read these drafts.
In large measure the progress we-have made has resulted from giving to the drafting group considerable latitude in using their k'nowledge and experience to reflect on the many valuable exchanges;with licensees and; 9-other interested parties and 'to incorporate within the draft specifics that resolve differences.
Our drafting group has spent many days in -
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discussion with groups from industry, professional societies, medical staffs, etc. to receive informal comments on the proposed revision.
Another facet of this effort has been that discussions with licensees and other outside groups have preceded any formal review within' the~ NRC; discussions within the NRC have also been informal.
This early input by i
licensees and other interested parties has permitted meaningful changes in the drafts before staff opinions on the provisions are effectively fixed by the formal in-house NRC staff reviews, which must take place before y
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. submittal to the-Commission sof a proposed rule,for publication in the Federal Register.
The response which we have received from'the licensees and others with' whom we had contact has been positive and constructive. Many issues have been considered and changes made to reduce what may be implementation difficulties; yet without sacrificing our objectives in initiating a revision or our responsibility to exercise independent judgement.
One such issue is the-dropping of the current 5(N-18) age - averaging dose formula along with the 3 rem quarterly d.ose limits for workers.
Under the present rule it is-
'Ii i I possible that a worker could receive 12 rems per year.
This would be HI' replaced by a 5 rem per year annual dose limit and a provision for " planned special exposures".
If a licensee establishes that it is necessary to perform specific important tasks which are likely to result in workers receiving doses in excess of the annual effective dose limits, the " planned special exposures" provision will provide a modest exposure " bank" for each enployee which can be used by the licensee to accomplish the tasks.
We know situations will arise during maintenance of nu' clear power systems in which exceeding the limits is warranted for continued operation and safety of the plant; and we provide for this when certain managerial conditions are met.
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_ not expect this provision to be conrnonly used and certainly not misused.
Let me specifically note that this requirement does not apply to emergency conditions.
Some additional issues wnich need further study have been identified and those are actively being pursued--in some cases as a joint effort between the industry and the NRC staff. We must, of course, get an accurate analysis of cost of implementing our revision.
How any revision of Part 20 will work is a question that we must work together on. Administrative procedures imposed by a revised Part 20 must be clear and consistently enforced. We must provide-practical regulatory guides that provide examples of how to meet the regulatory requirements.
Obviously to date we have only touched on some of these topics in developing our present draft; there is'much more we need to know about We are asking f' r that infonnation
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problems and the cost of implementation.
o at this stage because it will be more effective ~ now'than later.
I' encourage' you not to sit on the sidelines if you have substantive issues which you believe we should address.
What I am describing is, in our view, a dramatic and encouraging change in attitude betwe'en the NRC rulemakers and those being regulated or having interest in the proceedings. A few years ago under the initial leadership of Mr. Peter Libassi, then General Counsel for HEW, federal agencies' reviewed their radiation programs and reported their findings in a report dated June 1979.
In addition to recognizing the need for federal agencies to work together, the report noted "... that the decisionmaking process be opened further to public examination and participation... to' increase public confidence in federal agency activities." I believe this effort on Part 20 is a good example of meeting this recommendation and objective.
It is my
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hope tiis experience can become a trend because we can all benefit from it.
. As a fin'al renark on this issue let me make it. clear that while Ifi y
encouraged by the cooperation and constructive comments we have receivti, i
I am not suggesting that all. licensees have accepted this revision of Part-20 as a godsend--undoubtedly there remains some feeling of "why change?"
We recognize this reluctance, particularly because the level of protection provided (if that is the'only criterion to be used) is not. going to change.
dramatically.
I believe what will be accomplished by our efforts isLbetter assurance and understanding of.the protection provided, and a better conformance with the state of scientific knowledge and current practices. Because that assurance and und'ersta ing~ best serves the public interest I am optimistic I
l of a positive outcome lor bur: efforts.
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"De minimis" or Regulator'y Cut-off Values One of the features in the draft proposed revision of Part 20 is the establishment of a "de minimis" level or " regulatory cut-off" level for exposures to radiation.
It would constitute a level of risk so low that no' resources could be justified to control it, or - to be further concerned with it.
The need for a "de minimis" feature in the standards for protection.
against radiation has long been recognized in order to avoid over extending regulatory actions beyond what is needed to adequately protect public heal th. Applied to radiological protection, "de minimis" can be a level.
of risk (or dose rate, as a surrogate measure) so low that it would be trivial in comparison to the risks which the individual _is subjected to 3
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_ daily as part of normal living habits and activities.
It'is clear that
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public health protection does not require zero risk at-exorbitant costs-
-and therefore a regulatory cut-off point is'warrante'd.
I believe it is time to take.su~ch a step forward in our regulatory process and establish a level that, in effect, places a lower bound on the requireme'nt for considering what level of radiation protection constitutes ALARA.
I am pleased that Mr. Guy Cunningham of. NRC's legal staff will be i
providing you further thoughts on' this matter. He will, in 'particular, address the "de minimis" concept and its relationspip to specific-exceptions in regulations, such as any " exempt" quantities applicable to low-level ~ waste.
Such exceptions are warranted even.1f-the estimates of risk are-higher.than
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the "de minimis" values'when they 'a're'~bised on a case-by-caseev'alua' tion of.
ALARA considerations.
Because of the broad implications' of defining "de.minimis".or " regulatory cut-off" levels, I have proposed that this be a high priority issue for -
consideration by the Tnew Committee for Interagency Radiatio'n Protection..
Coordination, recently established by the Office of Science Technology and f
Policy. Cut-off-or "de minimis" values will be proposed in the draft revision of Part 20.
We expect considerable discussio~n on any numerical value selected.
l The important point is not so much the numerical values themselves but to initiate serious discussion of and find ' broad acceptance of the concept.
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During the course of this conference I am sure these and many other important radiation protection issues will be intensively discussed.
I look
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