ML20041E422

Testimony of Ef Branagan Re Contention 8/9 Re Dose Estimates in Fes.Prof Qualifications Encl

ML20041E422
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Site:	Waterford
Issue date:	03/08/1982
From:	Branagan E Office of Nuclear Reactor Regulation
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ISSUANCES-OL, NUDOCS 8203100503
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-# e UNITED STATES OF AMERICA NUCLEAR REGULATORY C0f2 FISSION 9EFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of

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LOUISIANA POWER AND LIGHT Docket No. 50-382 OL COMPANY

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Waterford Stean Electric Station,

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Unit 3)

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NRC STAFF TESTIM 0 fly OF ENIARD F. BRANAGAN, JR. REGARDING CONTENTION 8/9 0.1.

Please state your name and occupation.

A.I.

fly nane is Edward F. Branagan, Jr.

I am a Radiological Physicist with the Radiological Assessment Branch in the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission.

0.2.

Have you prepared a statenent of your professional cualifications?

A.2.

Yes. A copy is attached to this testimony.

Q.3.

Please describe the nature of your responsibilities.

A.3.

I an responsible for evaluating the environmental radiological l

impacts from nuclear power reactors and, in particular, for evaluating l

radiological models for use in reactor licensing.

1 Q.4.

What is the purpose of this testimony?

A.4.

This testinony is designed to address the dose estinates presented in the NRC Staff's Final Environnental Statement Related to the Operation of Waterford Steam Electric Station, Unit 3 (FES)

(NUREG-0779, September 1981), to the extent that those dose estimates 1

8203100503 820308 PDR ADOCK 05000382 T

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.* relate to Contention 8/9 in this proceeding. As utilized in this testinony, " dose" refers to the " dose equivalent" for an individual and the " dose commitment" for a population.

Q.5.

Have you reviewed Section 5.9.1 entitled " Radiological Impacts from Routine Operation" and Appendices H, I and J of the FES related to the operation of Waterford Unit 3?

A.S.

Yes. A copy of Section 5.9.1, as relevant, and Appendices H, I and J to the FES is attached to this testimony.

Q.6.

To the best of your knowledge and belief, are the statements -

set forth in Section 5.9.1 and Appendices H, I and J to the FES true and correct?

A.6.

Yes.

Q.7.

Has the NRC Staff calculated the amount of radioactive materials in liquid effluents to be released from Waterford Unit 3?

A.7.

Yes. The radioactive effluent releases calculated for the facilityarelistedinTableJ-8(p.J-10)oftheFES.

Q.8.

Has the Staff calculated doses to the public resulting from exposure to radioactive liquid releases?

A.8.

Yes. The radioactive liquid releases in Table J-8 (p. J-10) and the hydrological transport and dispersion factors in Table J-9 (p. J-11) of the FES were used to estimate doses (1) to a hypothetical maximally exposed individual, (?) to the population within 80 km of the plant, and (3) to the general U.S. population. These dose estimates are presented in Table J-5 (p. J-7) and Table J-7 (p. J-9) of the FES.

Q.9.

What is the Staff's estimate of the doses to a hypothetical maxinally exposed individual resulting from exposure to radioactive liquid releases?

in

.. A.9.

The total body dose to the maximally exposed individual resulting from exposure to radioactive liquid effluents from one year of reactor operation is about 0.1 nillirem (Table J-5).

This dose is a small fraction of the annual dose resulting fron exposure to natural background radiation (i.e., about 84 nillf rems for persons in the State of Louisiana).

Q.10. What is the Staff's estimate of the dose to the population resulting from exposure to radioactive liquid releases?

A.10.

The dose to the total body of the population within 80 km of the site resulting from exposure to radioactive liquid releases is about 6 person-rems (Table J-5).

This dose is a small fraction of the annual dose resulting fron exposure of the population to natural background radiation (i.e., about 180,000 person-rems within 80 km of the site).

Q.11. Has the Staff calculated the amount of radioactive materials in gaseous effluents to be released from Waterford Unit 37 A.11.

Yes.

The Staff's estinates of the quantities of radioactive gaseous effluents are presented in Table J-1 (p. J-4) of the FES.

The values in Table J-1 were used with the atmospheric dispersion factors in Table J-2 to estinate doses to a hypothetical maxinally exposed individual resulting from exposure to noble gases, and radiciodines and particulates.

The doses to the naximally exposed individual are l

presented in Tables J-4 and J-5.

Q.12. What is the Staff's estination of the dose to a hypothetical triximally exposed individual resulting fron exposure to radioactive gaseous releases?

A.12.

The dose to the total body of the maxinally exposed individual resulting from exposure to noble gases, or radioiodines and particulates, f

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se

, is estimated to be less than 5 millirems.

The highest dose to any organ (i.e., the bone of a child) was estimated to be about 12 millirems.

These doses are a fraction of the annual dose from exposure to natural background radiation (i.e., about 84 millirems for Louisiana).

Q.13.

What is the Staff's estimate of the dose to the population resulting from exposure to radioactive gaseous releases?

A.13.

The Staff estinated that the dose to the total body of the population within 80 kn of the site from exposure to radioactive gases (i.e., noble gases, radioiodines and particulates) would be about 6 person-rems (Table J-5).

This dose is a small fraction of the annual dose resulting from exposure of the population to natural background radiation (i.e., about 180,000 person-rens).

Q.14.

How do the doses to the hypothetical maximally exposed individual resulting from exposure to liquid and gaseous radioactive releases estinated by the Staff compare with the annual dose design objectives set forth in 10 C.F.R. Part 50 Appendix I?

A.14.

The Staff's dose estinates to the hypothetical maximally exposed i

individual resulting fron exposure to liquid and gaseous radioactive releases l

l from the facility are lower than the annual dose design objectives set forth

(

in 10 C.F.R. Part 50 Appendix I (see, e.g., FES Table J-5 (p. J-7)).

In addition, the dose design objectives set 'erth in 10 C.F.R. Part EJ Appen-dix I are about two orders of magnitene 5c av the dose limits for the public health and safety which can r; at,ad from 10 C.F.R. Part 20.

Q.15.

Please describe the environmental transport and dose nodels l

used by the Staff in estimating the doses referred to in your testimony.

l

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.. A.15.

In licensing connercial nuclear power reactors, the Staff uses mathematical models that characterize radionuclide movement in the environnent to determine the radiolocial impact resulting from nuclear pcwer plan + operations. These models are described in several NRC Regulatory Guides.

Regulatory Guide 1.109, entitled " Calculation of Annual Doses to ifan from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 C.F.R. Part 50, Appendix I,"

Revision 1 (October 1977), provides models for calculating doses to both the hypothetical maximally exposed individual and the general population resulting from exposure to radioactive liquid and airborne releases.

Other regulatory guides relating to this subject are Regulatory Guide 1.111,

" Methods for Estimating Atnospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Reactors", Revision 1 (July 1977); Regulatory Guide 1.112, " Calculation of Releases of Radio-active ffaterials in Gaseous and Liquid Effluents from Light-Water-Cooled rower Reactors," Revision 0-R (April 1976); and Regulatory Guide 1.113,

" Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," Revision 1 (April 1977).

The models employed by the NRC Staff were developed by members of the Staff having backgrounds and training in the field of radiological protection, and by experts at national laboratories such as Oak Ridge and Argonne under contract with the NRC. NRC models are sub.iect to l

continuing peer review and verification by other Federal agencies such as the Environmental Protection Ageiicy (EPA) and the Bureau of Radiological Health (BRH).

l

so EDWARD F. BRANAGAN, JR.

OFFICE OF NUCLEAR REACTOR REGULATION PROFESSIONAL OUALIFICATIONS From April 1979 to the present, I have been employed as a Radiolooical Physicist with the Radiological Assessment Branch in the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission (NRC). Currently, I am responsible for evaluating the environmental radiological impacts resulting from the operation of nuclear power reactors.

In particular, I am responsible for evaluating radio-ecological models and health effect models for use in reactor licensing.

In addition to my duties involving the evaluation of radiological impacts from nuclear reactors, my duties in the Radiological Assessment Branch have included the following:

(1) I managed and was the principal author of a report entitled " Staff Review of 'Radioecological Assessnent of the Wyhl Nuclear Power Plant'" (NUREG-0668); (2) I served as a tech-nical contact on an NRC contract with Argonne National Laboratory involv-ing development of a computer program to calculate health effects from radiation; (3) I served as the project manager on an NRC contract with Idaho National Engineering Laboratory involving estimated and measured concentrations of radionuclides in the environnent; (4) I served as the project manager on an NRC contract with Lawrence Livermore Laboratory concerning a literature review of values for parameters in terrestrial radionuclide transport models; and (5) I served as the project manager I

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-2 on an NRC contract with Dak Ridge National Laboratory concerning a statistical analysis of dose estimates via food pathways.

Fr n 1976 to April 1979, I was employed by the NRC's Office of Nuclear Materials Safety and Safeguards, where I was involved in project management and technical work.

I served as the project manager for the NRC in connection with the NRC's estimation of radiation doses from 1

radon-222 and radium-226 releases from uranium mills, in coordination with Dak Ridge National Laboratory which served as the NRC contractor.

As part of my work on NRC's Generic Environmental Impact Statement on Uranium Milling (GEIS), I estinated health effects fron uranium mill tailings. Upon publication of the GEIS, I presented a paper entitled

" Health Effects of Uraniun Mining and Milling for Commercial Nuclear Power" at a Conference on Health Implications of New Energy Technologies.

I received a B.A. in Physics from Catholic University in 1969, a M.A. in Science Teaching from Catholic University in 1970, and a Ph.D.

in Radiation Biophysics from Kansas University in 1976. While completing my course work for ny Ph.D., I was an instructor of Radiation Technology at Haskell Junior College in Lawrence, Kansas. My doctoral research work was in the area of DNA base damage, and was supported by a U.S.

Public Health Service traineeship; my doctoral dissertation was entitled

" Nuclear Magnetic Resonance Spectroscopy of Gamma-Irradiated DNA Bases."

I am a member of the Health Physics Society.

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NUREG-0779

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Fictal Environmental Statement re ated to the operation of Waterford Steam Electric Station, Unit No. 3 Docket No. 50-382 Louisiana Power and Light Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation September 1981 H

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5. 9 RADIOLOGICAL IMPACTS (1) Regulatory Requirements Nuclear power reactors in the United States must comply with certain regulatory requirements in order to operate.

The permissible levels of radiation in unrestricted areas and the radioactivity in effluents to unrestricted areas are spelled out in 10 CFR Part 20, Standards for Protection Against Radiation.38 These regulations specify limits on levels of radiation and limits on concen-trations of radionuclides in the station's effluent releases to the air and water (above natural background), under which the reactor must operate.

These regulations state that no member of the general public in unrestricted areas shall receive a radiation dose, due to Station operation, of more than 0.5 rems /yr (or 2 mrems/hr or 100 mrems/7 days) to the total body.

These radiation dose limits are established to be consistent with considerations of the health and safety of the public.

In addition to the Radiation Protection Standards of 10 CFR Part 20, there are spelled out in 10 CFR Part 50.36a license requirements that are to be imposed 39 on licensees in the form of Technical Specifications on Effluents from Nuclear Power Reactors to keep releases of radioactive materials to unrestricted areas during normal operations, including expected operational occurrences, as low as is reasonably achievable (ALARA).

Appendix I of 10 CFR Part 50 provides numerical guidance on design objectives and limiting conditions for operation of LWRs to meet this ALARA requirement.

Applicants for permits to construct and licenses to operate an LWR shall provide reasonable assurance that the following dose design objectives will be met:

3 mrems/yr to the total body or 10 mrems/yr to any organ from liquid effluents; 10 mrads/yr gamma radiation or 20 mrads/yr beta radiation from gaseous effluents--and/or 5 mrems/yr to the total body or 15 mrems/yr to the skin from gaseous effluents; and 15 mrems/yr to any organ from the airborne effluents that include the radiciodines, carbon-14, tritium, and the particulates.

Experience with the design, construction and operation of nuclear power reactors indicates that compliance with such design objectives will keep average annual releases of radioactive material in effluents at small percentages of the limits specified in 10 CFR Part 20, and in fact, generally below the design objective values of Appendix I.

At the same time, the licensee is permitted the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided a dependable source of power even under unusual operating conditions which may temporarily result in releases higher than such small percentages, but still well within the limits specified in 10 CFR Part 20.

5-27

y In addition to the impact created by station radioactive effluents as discussed above, within the NRC policy and procedures for environmental protection spelled out in 10 CFR Part 51 there are generic treatments of environmental effects of all aspects of the Uranium Fuel Cycle.

These environmental data have been sum-marized in Table S-3 (Table 5.13) and are discussed later in this report in Section 5.9.3 In the same manner the environmental impact of transportation of fuel and waste to and from an LWR is summarized in Table S-4 (Table 5.6) of Section 5.9.1.

Recently an additional operational requirement for Uranium Fuel Cycle Facilities including nuclear power plants has been established by the EPA in 40 CFR Part 190.40 This regulation limits annual doses (excluding radon and daughters) for members of the public to 25 mrems total body, 75 mrems thyroid, and 25 mrems other organs from all fuel cycle facility contributions that may impact a specific individuaTin the public.

(2) Operational Overview During normal operation of Waterford 3, small quantities of fission products and induced radioactivities will be released to the environment.

As required by NEPA, the staff has determined the dose estimated to members of the public outside of the plant boundaries due to the radiation from these radioisotope releases and relative to natural background radiation dose levels.

These station generated environmental dose levels are estimated to be very small due to plant design and the development of a conscious program which will be implemented at the station to contain and control all radioactive emissions and effluents.

As mentioned above, highly efficient radioactive-waste management systems are incorporated into the plant design and areThe specified in detail in the Technical Specifications for the station.

effectiveness of these systems will be measured by process and effluent radio-logical monitoring systems that permanently record the amounts of radioactive constitutents remaining in the various airborne and waterborne process and effluent streams.

The amounts of radioactivity released through vents and discharge points to be further dispersed and diluted to points outside the plant boundaries are to be recorded and published semiannually in the Radioactive Effluent Release Reports of each facility.

The small amounts of airborne effluents that are released will diffuse in the atmosphere in a fashion determined by the prevalent meteorological conditions and are thus much dispersed and diluted by the time they reach unrestricted areas that are open to the public.

Similarly, the small amounts of waterborne effluents released will be diluted with plant waste water and then further diluted as they are discharged into the Mississippi River beyond the plant boundaries.

Any radioisotopes in the station's effluents that finally enter unrestricted areas will produce dose effects through their radiations on members of the general public similar to the dose effects from background radiations (i.e.,

cosmic / terrestrial and internal radiations), which also include radiation from These radiation dose effects can be calculated for nuclear weapons fallout.

the many potential radiological exposure pathways specific to the environment around the station, such as direct radiation doses from the airborne or water-borne affluent streams outside of the plant boundaries, or internal radiation 5-28

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dose commitments from radioactive contaminants that might have been deposited on vegetation, or in meat and fish products eaten by people, or that might be present in drinking water outside the plant, or incorporated into milk from cows at nearby farms.

These doses, calculated for the " maximally exposed" individual (i.e., the hypothetical individual potentially subject to maximum exposure), form the basis of the NRC staff's evaluation of impacts.

These estimates are for a fictitious or " maximally exposed" person, since assumptions are made that tend to overestimate the dose that would actually accrue to members of the public outside the plant boundaries.

For example, if this " maximally exposed" indivi-dual were to receive the dose calculated at the plant boundary, he/she is assumed to be physically at that boundary for 100% of the year, and outside (unshielded from gamma radiation) 50% of the year, an unlikely occurrence.

Site specific values for the various parameters involved in each dose pathway are used in the calculations.

These include calculated or observed values for the amounts of radioisotopes released in the gaseous and liquid effluents, meteorological information (e.g., wind speed and directio.n) specific to the site topography and effluent release points, and hydrological information relative to dilu' tion and " flushing" of the liquid effluents as they are discharged.

A periodic land census, to be required by the Radiological Technical Specifications of the operating license, will require that as use of the land surrounding the site boundary changes, revised calculations be made to ensure that this dose estimate for gaseous effluents always represents the highest dose for any individual member of the public for each applicable foodchain pathway.

The estimate considers, for example, where people live, where vegetable gardens are located, where cows are pastured, etc.

For Waterford 3, in addition to the direct effluent monitoring, measurements will be made on a number of types of samples from the surrounding area to determine the possible presence of radioactive contaminants which, for l

example, might be deposited on vegetation, or be present in drinking water outside the plant, or incorporated into cow's milk from nearby farms.

l 5.9.1 Radiological Impacts from Routine Operations

5. 9.1.1 Radiation Exposure Pathways:

Dose Commitments There are many environmental pathways through which persons may be exposed to radiation originating in a nuclear power reactor.

All of the potentially meaningful exposure pathways are shown schematically in Figure 5.7.

When an individual is exposed via one of these pathways, his dose is determined in part by the amount of time he is in the vicinity of the source, or the amount of time the radioactivity is retained in his body.

The actual effect of the rad-iation or radioactivity is determined by calculating the dose commitment.

This dose commitment represents the total dose that would be received over a 50 yr period, following the intake of radioactivity for 1 yr under the conditions existing 15 yrs af ter the station begins operation (i.e., the mid point of station operation).

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P "WWDA Figure 5.7 Potentially Meaningful Exposure Pathways to Humans 5-30

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There are a number of possible expasure pathways to man that can be studied to determine whether the routine releases at the Waterford site are likely to have any significant impact on menhers of the general public living and working outside of the site boundaries, and whether the releases will in fact meet regulatory. requirements.

A detailed listing of these possibilities would include external radiation exposure from the gaseous effluents, inhalation of iodines and particulate contaminants in the air, drinking milk from a cow or eating meat from an animal that feeds on open pasture near the site on which iodines or particulates may have deposited, eating vegetables from a garden near the site that may be contaminated by similar deposits, drinking water and eating fish caught near the point of discharge of liquid effluents.

Other less significant pathways include:

external irradiation from radionuclides deposited on the ground surface, eating animals and food crops raised near the site using irrigation water that may contain liquid effluents, shoreline activities near lakes or streams that may be contaminated by effluents, and direct radiation from within the plant itself.

Calculations of the effects for most pathways are limited to a radius of 80 km (50 miles).

This limitation is based on several facts.

Experience has shown that all significant dose commitments (>0.1 mrems/yr) for radioactive effluents are accounted for within a radius of 80 km from the plant.

Beyond 80 km the doses are smaller than 0.1 mrems/yr, which is far below natural background doses, and the doses are subject to substantial uncertainty because of limitations of predictive mathematical models.

The NRC staff has made a detailed study of all of the above significant pathways and has evaluated the radiation dose commitments both to the plant workers and the general public for these pathways resulting from routine operation of the Station.

A di.cussion of these evaluations follows.

5.9.1.1.1 Occupational Radiation Exposure 5-31

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5.9.1.1.2 Public Radiation Exposure (1) Transportation of Radioactive Materials The transportation of " cold" (unitradiated) nuclear fuel to the reactor, of spent irradiated fuel from the reactor to a fuel reprocessing plant, and of solid radioactive wastes from the reactor to waste burial grounds is considered l

l in 10 CFR Section 51.20. " The contribution of the environmental effects of I

such transportation to the environmental costs of licensing the nuclear power reactor is set forth in Summary Table S-4 from 10 CFR Section 51.20, reproduced herein as Table 5.6.

The cumulative dose to the exposed population as summarized in Table S-4 is very small when compared to the annual dose of 26,000,000 person-rems to this same population from background radiation.

(2) Direct Radiation Radiation fields are produced around nuclear plants as a result of radioactivity within the reactor and its associated components, as well as a result of small radicactive effluent releases.

Direct radiation from sources within the plant are due primarily to nitrogen-16, a radionuclide produced in the reactor core.

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Table 5.5 Incidence of Job-Related Fatalities l

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Table 5.6 Environmental Impact of Transportation of Fuel and Waste To and From 1

One Light-Water-Cooled Nuclear Power Reactor NORMAL CONDITIONS OF TRANSPORT Environmental impact Heat (peri rradiated fuel cask in transit)............. 250,000 Btu /hr.

Weight (governed by Federal or State restrictions).... 73,000 lbs per truck; 100 tons per cask per rail car.

Traffic density:

Truck......

.........................................Less than 1 per day.

R a i l............................................... L e s s tha n 3 pe r mo nth.

Exposed population Estimated Range of doses to Cumulative dos'

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number of exposed individuals exposed population persans (per reactor year)

(per reactor year)3 exposed Transportation workers..

200......

0.01 to 300 millirem...

4 person-rem General public:

Onlookers.............

1,100......

0.003 to 1.3 millirem..

3 person-rem Along Route...........

600,000......

0.0001 to 0.06 millirem.

ACCIDENTS IN TRANSPORT Environmental risk Radiological effects...................Small4 Common (nonradiological causes).......1 fatal injury in 100 reactor years; 1 nonfatal injury in 10 reactor years;

$475 property damage per reactor year.

2 Data supporting this table are given in the Commission's " Environmental Survey of Transportation of Radioactive Materials to and from Nuclear Power Plants," WASH-1238, December 1972, and Supp. 1, NUREG-75/038, April 1975.

Both documents are available for inspection and copying at the Commission's Public Document Room, 1717 H Street, NW.,

Washington, D.C., and may be obtained from National Technical Information Service, Springfield, VA 22161.

WASH-1238 is available from NTIS at a cost of $5.45 (micro-fiche, $2.25) and NUREG-75/038 is available at a cost of $3.25 (microfiche, $2.25).

2The Fedeial Radiation Council has recommended that the radiation doses from all sources of radiation other than natural background and medical exposures should be limited to 5,000 millirem per year for individuals as a result of occupational enposure and should be limited to 500 millirem per year for individuals in the general population.

The dose to individuals due to average natural background radiation is about 130 millirem per year.

3 Person-rem is an expression for the summation of whole body doses to individuals i

in a group.

Thus, if each member of a population group of 1,000 people were to receive a dose of 0.001 rem (1 millirem), or if 2 people were to receive a dose of 0.5 rem (500 millirem) each, the total person-rem dose in each case would be 1 person-rem.

4Although the environmental risk of radiological effects stemming from trans-portation accidents is currently incapable of being numerically quantified, the

.-isk remains small regardless of whether it is being applied to a single reactor or a multireactor site.

1 5-34 l

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Because the primary coolant of a PWR is contained in a heavily shielded area, dose rates in the vicinity of PWRs are generally undetectable (less than 5 mrems/yr).

Low-level radioactivity storage containers outside the )lant are estimated to make a dose contribution at the site boundary of less tlan 0.1% of that due to the direct radiation described above.

(3) Radioactive Effluent Releases:

Air and Water As pointed out in section 4.2.3, all effluents from the station will be subject to extensive decontamination, but small controlled quantities of radioactive effluents will be released to the atmosphere and to the hydrosphere during normal operations.

Estimates of site specific radioisotope release values have been developed on the basis of the description of operational and radwaste systems in the applicant's ER and FSAR and by using the calculational model and parameters developed in NUREG-0017.45 This has been supplemented by extensive use of the applicant's site and environmental data in the ER and in subsequent answers to NRC staff questions, to obtain a complete picture of airborne and waterborne releases from the station.

These small amounts of effluents are then highly diluted by the air and water into which they are released before the with activities of the general public. y reach areas in which they interact Radioactive effluents can be divided into several groups.

Amon effluents the radioisotopes of the noble gases--krypton, xenon,g the airborne and argon--do not deposit on the ground or interact with living organisms; therefore the noble gas ef fluents act primarily as a source of direct external rn. a,. ion emanating from the effluent plume.

Dose calculations are performed for the site boundary where the highest external radiation doses to a member of the public as a result of gaseous effluents have been estimated to occur; these include the annual beta and gamma air doses as well as the total body and skin doses from the plume at that boundary location.

Another group of airborne radioactive effluents--the radioiodines, carbon-14, and tritium--are also gaseous but tend to be deposited on the ground and/or absorbed into the body during inhalation.

For this class of effluents, esti-mates of direct external radiation doses from deposits on the ground, and of internal radiation doses to total body, thyroid, bone, and other organs from inhalation, from vegetable consumption, from milk consumption, and from meat consumption are made.

Concentrations of iodine in the thyroid and of carbon-14 in bone are of particular significance here.

A third group of airborne effluents, consisting of particulates that remain after filtration of the effluents could include fission products such as cesium and barium and corrosion pr,oducts such as cobalt and chromium.

The calculational model determines the direct external radiation dose and the internal radiation doses for these contaminants through the same pathways as described above for the radiciodines, carbon-14, and tritium.

Doses from the particulates are combined with those of the radioiodines, carbon-14, and tritium for comparison to one of the design objectives of Appendix I to 10 CFR Part 50.

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The waterborne radioactive effluent constituents could include fission products corrosion and activation products, such as such as strontium and iodine; tium as tritiated water.

Calculations estimate and tri sodium and manganese;f any) from fish consumption, from water ingestion (as the internal doses (i drinking water), and from eating of meat or vegetables raised near the site on irrigation water, as well as any direct external radiation from recreational use of the water past the point of discharge.

The release values for each group of effluents along with site-specific meteoro-logical and hydrological data, serve as input to computerized radiation-dose models that estimate the maximum radiation dose that would be received outside the facility via a number of pathways for individual members of the public and for the general public as a w1 ole.

These models and the radiation dose calcula-tions are discussed in Regulatory Guide 1.10f86 and in Appendix H of this statement.

Examples of site-specific dose assessment calculations and discussions of Doses from all airborne effluents parameters involved are given in Appendix J.

except the noble gases are calculated for the location (e.g., site boundary, garden, residence, milk cow, meat animal) where the highest radiation dose to a member of the public from all applicable pathways has been established.

Only those pathways associated with airborne effluents that are known to exist at a single location, are combined to calculate the total maximum exposure to an exposed individual.

Pathways associated with liquid effluents are combined without regard to location, but they are assumed to be associated with maximum exposure to an individual other than through gaseous-effluent pathways.

5.9.1.2 Radiological Impact on Humans Although the doses calculated in Appendix J are based on radioactive-waste treatment system capability, the actual radiological impact associated with the operation of the station will depend, in part, on the manner in which the radioactive waste treatment system is operated.

Based on its evaluation of the potential ?erformance of the ventilation and radwaste treatment systems, the NRC staf f 1as concluded that the systems as now pro)osed are capable of controlling effluent releases to meet the dose design o)jectives of Appendix I to 10 CFR Part 50.39 lhe Station's operation will be governed by, operating license Technical Specifi-cations which will be based on the dose design objectives of Appendix I to Since these design objective values were chosen to permit 10 CFR Part 50.39 flexibility of operation while still ensuring that plant operations are ALARA, the actual radiological impact of plant operation may result in doses close to the dose design objectives.

Even if this situation exists, the individual doses for the member of the public subject to maximum exposure will still be very small when compared to natural background doses ($100 mrems/yr) or the dose limits specified in 10 CFR Part 20 (500 mrems/yr - whole body).

As a result, the staff concluded that there will be no measurable radiological impact on members of the public from routine operation of the station.

's Operatingstandardsof40CFRPart190,theEnvironmentalProtectionAgencyO Environmental Radiation Protection Standards for Nuclear Power Operations, specify that the annual dose equivalent must not exceed 25 mrems to the whole 5-36

body, 75 mrems to the thyroid, and 25 mrems to any other organ of any member materials (radon and its daughters excepted) planned discharges of rad of the public as the result of exposures to to the general environment from all uranism-fuel-cycle operations and radiation from these operations that can be expected to affect a given individual.

The NRC staff concludes that under normal operations Waterford 3 is capable of operating within these standards.

The radiological etfects of a nuclear power plant are well known and documented.

Accurate measurements of radiation and radioactive contaminants can be made with very high sensitivity so that much smaller amounts of radioisotopes can be recorded than can be associated with any possible known ill effects.

Furthermore, the effects of radiation on living systems have for decades been subjecttointensiveinvestigationandconsiderationbyindividualscientists as well as by select committees, occasionally constituted to objectively and independently assess radiation dose effects.

Although, as in the case of chemical contaminants, there is debate about the exact extent of the effects of very low levels of radiation, the limits of deleterious effects are well established and amenable to standard methods of risk analysis.

Thus the risks to the maximally exposed member of the public outside of the site boundaries can be readily quantified.

Further, the impacts on, and risks to, the total population outside of the boundaries can also be readily calculated and recorded.

5.9.1.3 Radiological Impacts on Biota Other Than Humans Depending on the pathway and radiation source, terrestrial and aguatic biota will receive doses that are approximately the same or somewhat higher than humans receive.

Although guidelines have not been established for acceptable limits for radiation exposure to species other than human, it is generally agreed that the limits established for humans are conservative fer other species.

Experience has shown that it is the maintenance of population stability that is crucial to the survival of a species, and species in'most ecosystems suffer rather high mortality rates from natural causes.

While the existence of extremely radiosensitive biota is possible, and while increased radiosensitivity in organisms may result from environmental inter-actions with other stresses (for example, heat or biocides), no biota have yet been discovered that show a sensitivity (in terms of increased morbidity or I

mortality) to radiation exposures as low as those expected in the area surrounding the station.

Furthermore, at all nuclear plants for which radia-tion exposure to biota other than humans has been analyzed,47 there have been no cases of exposure that can be considered significant in terms of harm to the species, or that approach the limits for exposure to members of the public 48 concluded that that are permitted by 10 CFR Part 20.38 The 1972 BEIR Report the evidence to date indicates that no other living organisms are very much more radiosensitive than humans; therefore, no measurable radiological impact on populations of biota is expected as a result of the routine operation of l

this station.

5.9.1.4 Radiological Monitoring Radiological environmental monitoring programs are established to provide' data on measu able levels of radiation and radioactive materials in the site 5-37 l

i

environs.

Such monitoring programs are conducted to verify the effectiveness of in plant systems used to control the release of radioactive materials and to ensure that unanticipated buildups of radioactivity will not occur in the environment.

Secondarily, the monitoring programs could identify the highly unlikely exis,tence of previously undetected releases of radioactivity.

A surveillance (Land Census) program is established to identify changes in the use of unrestricted areas to provide a basis for modifications of the monitoring programs.

These programs are discussed in greater detail in NRC Regulatory Guide 4.1, Rev.1, " Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants,"

and the Radiological Assessment Branch Technical Position, Rev. 1 November 1979, "An Acceptable Radiological Environmental Monitoring Program.A50 5.9.1.4.1 Preoperational The preoperational phase of the monitoring program should provide for the measurement of background levels of radioactivity and radiation and their var-iations along the anticipated important aathways in the areas surrounding the station, the training of personnel and tie evaluation of procedures, equipment and techniques.

The ap)licant proposed a radiological environmental-monitoring programtomeettheseo]jectivesintheER-CPanditwasdiscussedintheFES-CP.

This early program has been updated and expanded; it is presented in Section 6.1.5 of the applicant's ER-0L and is summarized here in Table 5.7.

The applicant states that the preoperational program has been implemented, at least two years prior to initial criticality of Waterford 3, to document back-ground levels of direct radiation and concentrations of radionuclides that exist in the environment.

The preoperational program will continue up to the initial criticality of Waterford 3 at which time the operational radiological monitoring program will commence.

The staff has reviewed the preoperational environmental monitoring plan of the applicant and finds that it is acceptable as presented.

5.9.1.4.2 Operational The operational, offsite radiological-monitoring program is conducted to measure radiation levels and radioactivity in plant environs.

It assists and provides backup support to the ef fluent-monitoring program as recommended in NRC Regulator)

Guide 1.21, " Measuring, Evaluating and Re)orting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants."51 The applicant states that the operational program will in essence be a continua-tionofthepreoperationalprogramdescribedabovewithsomeadjustmentof sampling frequencies in expected critical exposure pathways, such as increasing milk sampling frequency and deletion of fruit, vegetable, soil, and gamma radiation survey samples.

The proposed operational program will be reviewed prior to plant operation.

Moditication will be based upon anomalies and/or exposure pathway variations observed during the preoperational program.

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Table 5.7 Radiological Environmental Monitoring Program for Waterford Fxposure pathway Number of samples

• Sampling and Type of frequency and/or sample type and locations collection frequency of analysis AIRBORNE Radiofodine and __

3 offsite locations (in Continuous sampler operation Radiciodine cartridge:

particulates different sectors) of with sample collection Analyze weekly for 1-131 the highest calculated weekly or as required by annual average ground dust loading, whichever Particulate sampler:

level D/Q (A8-A16-S,A17-NW)g, is more frequent Gross beta radioactivity following filter change, composite (by location)

I sample from the vicin-for gamma isotopic Ity of Killona, a com-quarterly munity having the highest calculated annual average ground level D/Q.

I sample from the vicin-ity of Norco (A13) and I sample from Laplace (A14) 1 sample from Desa11emond (A12. SSE) 1 sample from Luling (All-E),

a control location 10-20 miles distant and in a leastpregalentwind direction O! RECT RADIAT!DN TLD 4 stations at s500 ft Quarterly, semi-annually Gamma dose quarterly in W, WNW, 5, and NW sectors.

8 stations 1 mile from plant in SSE, S. SSW, W5W, W, NW, N, and NE sectors Norco (W)

Laplace Luling (E)

Desallemond (SSE) 4 stations located in special interest areas.

WATERBORNE Surface" 1upst5*****('2 C "P*' ***PA'

'""* ' pic '"'

miles) 1 month period monthly. Composite for 1 downstream sample tritium analysis quarterly

($1000 meters) 1 sample from intake structure Ground Riverside of plant Quarterly Gamma isotopic and tritium (G1) analysis quarterly Lakeside of plant (G2)

Drinking 1 sample from Union Monthly composite taken at Gross beta and gamma isotopic Carbide (W7) each municipal facility isotopic analysis monthly.

1 sample from Composition for tritium St. Charles Parish (W8) analysis quarterly Rooted aquatic I sample 1000 meters Semiannual Gamma isotopic analysis plants & shore-downstream semiannually line sediments 1 sample 2 miles upstream Bottom sediments 1 sample 1000 meters Semiannual Gamma isotopic analysis downstream semiannually 1 sample 2 miles upstream 5-39

e,.

Table 5.7 Continued bposure pathway fumber of samples Sampling and Type of frequency a

and/or sample type and locations collection frequency of analysis

]NGE5110N Fish and 1 sample 1000 meters Semiannual Gamma isotopic analysis of invertebrates

- downstream edible portions I sample 2 mi upstream k

Fruits and vegetables Samples from following At time of harvest Gamma isotopic analysis of edible locations portions 1 mile NW (A15) 1 mile NE (A19) 1 mile N (A20) 1.7 mile N (A20) 1.3 mile W(A21)

Luling Desallemond Milk Samples from following Semimonthly when animals Gamma isotopic and 1-131 locations:

are on pasture, monthly at analyses semimonthly when 1 mile NW (A15) other times animals are on pasture.

1.7 mile N (A20) monthly otherwise

• 1.3 mile W (A21)

Luling Desallemond Meat animals Samples from following Semiannually for wildlife.

Gamma isotopic analysis on

~

edible sections semiannually locations:

1 mile NE (A19)

Luling Desallemond "The number, media, frequency, and location of samples may vary. It is recognized that, at times, it may not In be possible or practical to obtain samples of the media of choice at the most desired location or time.

these instances suitable alternative media and locations may be chosen for the particular pathway in ques-tion and submitted for acceptance.

The parenthetical symbols correspond to the location identification specified in figures 6.1.5-2 and 6.1.5-3 of the applicant's Environmental Report.

Particulate sample filters are analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more af ter sampling to allow C

for raden and thoron daughter decay, if gross beta activity in air or water is greater than ten times the yearly mean of control samples for any medium, gamma isotopic analysis will be performed on the individual

samples, Gamma isotopic analysis means the identification and quantification of gamma emitting radionuclides that may d

be attributable to the effluents from the facility.

'The purpose of this sample is to obtain background information.

Regulatory Guide 4.13 provides minimum acceptable performance criteria for TLD systems used for environc. ental monitoring. One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate For the purpose of this continuously, may be used in place of, or in addition to, integrating dosimeters.

table, a thermoluminescent dosimeter may be considered to be one phosphorus and two or more n50 sors in a packet cay be considered as two or more dosisters. The 40 stations are not an absolute number.

The "down-The " upstream sample" will be taken at a distance beyond significant influence of the discharge.

9 stream" sample will be taken in an area beyond but near the mixing zone.

Com osite samples will be collected with equipment (or equivalent) which is capable of collecting an aliquot h

at time intervals which are very short (e.g., hourly) relative to the compositing period (e.g., monthly).

' Groundwater samples will be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.

3The dose will be calculated for the maximum organ and age group, using the methodology contained in Regulatory Guide 1.109, and the actual parameters particular to the site.

If harvest If harvest occurs more than once a year, sampling will be performed during each discrete harvest.

occurs continuously, sampling will be monthly. Attention will be paid to including samples of tuborous and root food products.

5-40

The final operational-monitoring program proposed by the applicant will be reviewed in detail by the NRC staff, and the specifics of the required monitoring program will be incorporated into the Operating License Radiological Technical Specifications.

f l

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5-41

APPENDIX H NEPA POPULATION-DOSE ASSESSMENT 1

f 9

H-1

Population-dose commitments are calculated for all individuals living within 80 km (50 miles) of Waterford 3, employing the same models used for individual doses (see Regulatory Guide 1.109, Rev. 1)1, for the purpose of meeting the "as low as reasonably achievable" (ALARA) requirements of 10 CFR, Part 50, Appendix I.2__In addition, dose commitments to the population residing beyond the 80-km region, associated with the export of food crops produced within the 80-km region and with the atmospheric and hydrospheric transport of the more mobile effluent species, such as noble gases, tritium, and carbon-14, are taken into consideration for the purpose of meeting the requirements of the National Environmental Policy Act, 1969 (NEPA).

This appendix describes the methods used to inake these NEPA population dose estimates.

1.

Iodines and Particulates Released to the Atmosphere Effluent nuclides in this category deposit on the ground as the effluent moves downwind, thus the concentration of these nuclides remaining in the plume is continuously being reduced.

Within 80 km of the facility, the deposition model in Regulatory Guide 1.111, Rev.1,3 is used in conjunction with the dose models in Regulatory Guide 1.109, Rev. 1.1 Site specific data concerning production and consumption of foods within 80 km of the reactor are used.

For estimates of population doses beyond 80 km it is assumed that excess food not consumed.

within the 80-km area will be consumed by the population beyond 80 km.

It is further assumed that none, or very few, of the particulates released from the facility will be transported beyond the 80-km distance; thus they will make no contribution to the population dose outside the 80-km region.

This assumption was tested and found to be reasonable for Waterford 3.

2.

Noble Gases, Carbon-14, and Tritium Released to the Atmosphere For locations within 80 km (50 miles) of the reactor f acility, exposures to these effluents are calculated with a constant mean wind-direction model according to the guidance provided in Regulatory Guide 1.111, Rev. 1, and the dose models described in Regulatory Guide 1.109, Rev.1.

For estimating the dose commitment from these radionuclides to the U.S. population residing beyond the 80-km region, two dispersion regimes are considered.

These are referred to as the first pass dispersion regime and the world-wide dispersion regime.

l The model for the first pass dispersion regime estimates the dose commitment to the population from the radioactive plume as it leaves the facility and drifts across the continental United States to the northeastern corner of the U.S.

The model for the world-wide dispersion regime estimates the dose commitment to the U.S. population after the released radionuclides mix uniformly in the world's atmosphere or oceans.

a.

First-Pass Dispersion For estimating the dose commitment to the U.S. population residing beyond the 80-km region due to the first pass of radioactive pollutants, it is assumed that the pollutants disperse in the lateral and vertical directions along the plume path.

The direction of movement of the plume is assumed to be from the facility toward the northeast corner of the U.S.

The extent of vertical dispersion is assumed to be limited by the ground plane and the stable atmospheric layer aloft, the height of which determines the mixing depth.

The H-2

shape of such a plume geometry can be visualized as a right cylindrical wedge whose height is equal to the mixing depth.

Under the assumption of constant population density, the population dose associated with such a plume geometry is independent of the extent of lateral dispersion, and is only dependent upon the mixing depth and other nongeometrical related factors.4 The mixing depth is esti-2 mated to be 1000m, and n uniform population density of 62 persons /km is assumed along the plume path, with an average plume transport velocity of 2 m/s.

The total-body population dose commitment from the first pass of radioactive effluents is due principally to external exposure from gamma-emitting noble gases, and to internal exposure from inhalation of air containing tritium and from ingestion of food containing carbon-14 and tritium, b.

World-Wide Dispersion For estimating the dose commitment to the U.S. population after the first pass, world-wide dispersion is assumed.

Nondepositing radio-nuclides with half-lives greater than one year are considered.

Noble gases and carbon-14 are assumed to mix uniformly in the world's 3

atmosphere (3.8 x 1018 m ), and radioactive decay is taken into con-sideration.

The world-wide dispersion model estimates the activity of each nuclide at the end of a 15 year release period (midpoint of reactor life) and estimates the annual population dose commitment at that point in time, taking into consideration radioactive decay.

The total-body population dose commitment from the noble gases is due mainly to external exposure from gamma-emitting nuclides, while from carbon-14 it is due mainly to internal exposure from ingestion of food containing carbon-14.

The population dose commitment due to tritium releases is estimated in a manner similar to that for carbon-14, except that after the first-pass, all of the tritium is assumed to be absorbcd by the world's oceans (2.7 x 10" m ).

The concentration of tritium in the world's 3

oceans is estimated at the point in time after 15 years of releases have occurred, taking into consideration radioactive decay; the population dose commitment estimates are based on the incremental concentration at that point in time.

The total-body population dose commitment from tritium is due mainly to internal exposure from the l

consumption of food grown with irrigation water.

3.

Liquid Effluents Population dose commitments due to effluents in the receiving water within 80 km (50 miles) of the facility are calculated as described in Regulatory Guide 1.109.

It is assumed that no depletion by sedimentation of the nuclides present in the receiving water occurs within 80 km.

It also is assumed that aquatic biota concentrate radioactivity in the same manner as was assumed for the ALARA maximumally exposed individual evaluation.

However, food consumption values appropriate for the average, rather than the maximum, individual are used.

It is further assumed that all the sport and commercial fish and shellfish caught within the 80-km area are eaten by the U.S. population.

H-3

• s Beyond 80 km, it is assumed that all the liquid-effluent nuclides except tritium have deposited on the sediments so that they make no further contribution to population exposures.

The tritium is assumed to mix uniformly in the hydrosphere and to result in an exposure to the U.S. population in the same manner as discussed for -tritium in gaseous effluents.

REFERENCES FOR APPENDIX H (1)

U.S. Nuclear Regulatory Commission, " Regulatory Guide 1.109:

Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977.

(2) Title 10 Code of Federal Regulations Part 50, " Domestic Licensing of Production and Utilization Facilities," January 1981.

(3)

U.S. Nuclear Regulatory Commission, " Regulatory Guide 1.111:

Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Reactors," Revision 1, July 1977.

(4)

K. F. Eckerman, et. al., " User's Guide to JASPAR Code," U.S. Nuclear Regulatory Commission report NUREG-0597, Jtne 1980.

H-4

APPENDIX I IMPAC7 0F THE URANIUM FUEL CYCLE I-1

The following assessment of the environmental impacts of the fuel cycle as related to the operation of the proposed project is based on the values given in Table S-3 (Section 5.9) and the NRC staff's analysis of the radiological impact from radon releases.

For the sake of consistency, the analysis of fuel cycle impacts tas been cast in terms of a model 1000-MWe light-water-cooled reactor (LWR) operating at an annual capacity factor of 80%.

In the following review and evaluation of the environmental impacts of the fuel cycle, the staff's analysis and conclusions would not be altered if the analysis were to be based on the net electrical power output of Waterford 3.

1.

Land Use The total annual land requirement for the fuel cycle supporting a model 2

2 1000-We LWR is about 460,000 m (113 acres).

Approximately 53,000 m (13 acres) 2 per year are permanently committed land, and 405,000 m (100 acres) per year are temporarily committed.

(A " temporary" land commitment is a commitment for the life of the specific fuel cycle plant, e.g., mill, enrichment plant, or succeeding plants.

On abandonment or decommissioning, such land can be used for any purpose.

" Permanent" commitments represent land that may not be released for use after plant shutdown and/or decommissioning.) Of the 2

2 are undisturbed 405,000 m per year of temporarily committed land, 320,000 m 2 are disturbed.

Considering common classes of land use in the and 90,000 m United States,* fuel cycle land use requirements to support the model 1000-MWe LWR do not represent a significant impact.

2.

Water Use The principal water use requirement for the fuel cycle supporting a model 1000-MWe LWR is that required to remove waste heat from the power stations supplying electrical energy to the enrichment step of this cycle.

Of the total annual requirement of 43 x 10s m (11.4 x 109 gal), about 42 x 106 3

3 m are required for this purpose, assuming that these plants use once-through cooling.

Other water uses involve the discharge to air (e.g., evaporation losses in process cooling) of about 0.6 x 106 ma (16 x 107 gal) per year and water discharged to the ground (e.g., mine drainage) of about 0.5 x 106 a

m per year.

On a thermal offluent basis, annual discharges from the nuclear fuel cycle are about 4% of tt.e model 1000-MWe LWR using once-through cooling.

The consumptive water use of 0.6 x 106 3 per year is about 2% of the model 1000-MWe LWR using m

cooling towers.

The maximum consumptive water use (assuming that all plants supplying electrical energy to the nuclear fuel cycle used cooling towers) would be about 6% of the model 1000-MWe LWR using cooling towers.

Under this condition, thermal effluents would be negligible.

The staff finds that these combinations of thermal loadings and water consumption are acceptable relative to the water use and thermal discharges of the station.

mA coal-fired plant of 1000-MWe capacity using strip-mined coal requires the 2

disturbance of about 810,000 m (200 acres) per year for fuel alone.

I-2

3.

Fossil Fuel Consumption Electrical energy and process heat are required during various phases of the fuel cycle process.

The electrical energy is usually produced by the combustion of fossil fuel at conventional power plants.

Electrical energy associated with the fuel tycle represents about 5% of the annual electrical power production of the model 1000-MWe LWR.

Process heat is primarily generated by the combus-tion of natural gas.

This gas consumption, if used to generate electricity, would be less than 0.3% of the electrical output from the model plant.

The staff finds that the direct and indirect consumptions of electrical energy for fuel cycle operations are small and acceptable relative to the net power production of the station.

4.

Chemical Effluents The quantities of chemical, gaseous, and particulate effluents associated with fuel cycle processes are given in Table S-3.

The principal species are 50x NO, and the particulates.

Judging from data in a Council on Environmental Quflity report,1 the NRC staff finds that these emissions constitute an extremely small additional atmospheric loading in comparison with these emis-sions from the stationery fuel-combustion and transportation sectors in the United States, that is, about 0.02% of the annual national releases for each of these species.

The staff believes such small increases in releases of these pollutants are acceptable.

l Liquid chemical effluents produced in fuel cycle processes are related to fuel enrichment, fabrication, and reprocessing operations and may be released to receiving waters.

These effluents are usually present in dilute concentrations such that only small amounts of dilution water are required to reach levels of l

concentration that are within established standards.

Table S-3 specifies the flow of dilution water required for specific constituents.

Additionally, all liquid discharges into the navigable waters of the United States from plants associated with the fuel cycle operations will be subject to requirements and l

limitations set forth in the NPDES permit.

l Tailings solutions and solids are generated during the milling process.

These solutions and solids are not released in quantities sufficient to have a significant impact on the environment.

5.

Radioactive Effluents l

Radioactive effluents estimated to be released to the environment from l

reprocessing and waste management activities and certain other phases of the l

fuel cycle process are set forth in Table S-3.

Using these data, the staff has calculated the 100 year involuntary environmental dose commitment

• to the U.S. population.

l These calculations estimate that the overall involuntary total-body gaseous dose commitment to the U.S. population from the fuel cycle (excluding reactor releases and the dose commitment due to radon-222) would be approximately A

The environmental dose commitment (EDC) is the integrated population dose for 100 years; that is, it erepresents the sum of the annual population doses for a total of 100 years.

The population dose varies with time, and it is not practical to calculate this dose for every year.

I-3

400 person-rems per year of operation of the model 1000-K4e LWR.

Based on Table S-3 values, the additional involuntary total body-dose commitments to the U.S. population from radioactive liquid effluents due to all fuel cycle operations other than reactor operation would be approximately 100 person-rems per year of operation.

Thus the estimated involuntary 100 year environmental dose commitment to the U.S. population from radioactive gaseous and liquid releases due to these portions of the fuel cycle is approximately 500 person-rems (whole-body) per year of operation of the model 1000-MWe LWR.

At this time Table S-3 does not address the radiological impacts associated with radon-222 releases.

Principal radon releases occur during mining and milling operations and as emissions from mill tailings.

The staff has deter-mined that releases from these operations for each year of operation of the model 1000-Kde LWR are as given in Table I-1.

The staff has calculated population dose commitments for these sources of radon-222 using the RABGAD computer code described in Appendix A of Chap. IV, Sec. J, of NUREG-002.2 The results of these calculations for mining and milling activities prior to tailings stabilization are listed in Table I-2.

When added to the 500 person-rems total-body dose commitment for the balance of the fuel cycle, the overall estimated total-body involuntary 100 year environmental dose commitment to the U.S. population from the fuel cycle for the model 1000-Rde LWR is approximately 640 person-rems.

Over this period of dose is equivalent to 0.00002% of the natural background dose of about 3 billion person rems to the U.S. population.*

The staff has considered the health effects associated with the releases of radon-222, including both the short-term effects of mining and milling, and active tailings, and the potential long-term effects from unreclaimed open pit mines and stabilized tailings.

The staff has assumed that after completion of active mining underground mines will be sealed, returning releases of radon-222 to background levels.

For purposes of providing an upper-bound impact assess-ment, the staff has assumed that open pit mines will be unreclaimed and has calculated that if all ore were produced from open pit mines, releases from them would be 110 Ci per year per reference reactor year (RRY).

However, because the distribution of uranium ore reserves available by conventional mining methods is 66.8% underground and 32.2% open pit,3 the staff has further assumed that uranium to fuel LWRs will be produced by conventional mining methods in these proportions.

This means that long-term releases from unreclaimed open pit mines will be 0.332 x 110 or 37 Ci per year per RRY.

ABased on an annual average natural background individual dose commitment of 100 millirems and a stabilized U.S. population of 300 million.

l I

• s Table I-1 Radon releases for each year of operation of the model 1000-MWe LWR *

-Radon source Quantity released Source Mining 4060 Ci a

Milling and tailings (during active mining) 780 Ci b

Inactive tailings (prior to stabilization) 350 Ci b

Stabilized tailings (several hundred years)

I to 10 Ci/ year b

Stabilized tailings (after several hundred years) 110 Ci/ year b

aR. Wilde, U.S. Nuclear Regulatory Commission transcript of direct testimony given "In the Matter of Duke Power Company Company (Perkins Nuclear Station), Docket No. 50-488, April 17, 1978.

bP. Magno, U.S. Nuclear Regulatory Commission transcript of direct testimony given "In the Matter of Duke Power Company (Perkins Nuclear Station)," Docket No. 50-448, April 17,1978.

• After three days of hearings before the Atomic Safety and Licensing Appeal Board (ASLAB) using the Perkins record in a

" lead case" approach, the ASLAB issued a decision on May 13, 1981 (ALAB-640) on the radon-222 release source term for the Uranium Fuel Cycle.

The decision, among other matters, produced new source term numbers based on the record developed at the hearings.

These new numbers did not differ significantly from those in the Perkins record which are the values set forth in this Table.

Any health effects relative to radon-222 are still under consideration before the A5 LAB.

Since the source term numbers in ALAB-640 do not differ significantly from those in the Perkins record, the staff continues to conclude that "both the dose commitments and health effects of the uranium fuel cycle are insignificant when compared to dose commitments and potential health effects to the U.S. population resulting from all natural background sources." (see page I-7)

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• e

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Table 1-2 Estimated 100 year environmental dose commitment per year of operation of the model 1000-MWe LWR Dosage (person-rems)

Radon Source Releases (Ci)

Total Body Bone Lung (Bronchial epithelium)

Mining 4100 110 2800 2300 Milling and active tailings 1100 29 750 620 Total 140 3600 2900 Based on the above, the radon released from unreclaimed open pit mines over 100- and 1000 year periods would be about 3700 Ci and 37,000 Ci per RRY respectively.

The total dose commitments for a 100 to 1000 year period would be as follows:

Population dose commitments (person-rems)

Time span (years)

Releases (Ci)

Total Bone Lung (bronchial body epithelium) 100 3,700 96 2,500 2,000 500 19,000 480 13,000 11,000 1,000 37,000 960 25,000 20,000 The above dose commitments represent a worst-case situation in that no mitigating circumstances are assumed.

However, state and Federal laws currently require reclamation of strip and open pit coal mines, and it is very probable that similar reclamation will be required for uranium open pit mines.

If so, long-term releases from such mines should approach background levels.

For long-term radon releases from stabilized tailings piles, the staff has assumed that these tailings would emit, per RRY,1 Ci per year for 100 years, 10 Ci per year for the next 400 years and 100 Ci per year for periods beyond With these assumptions, the cumulative radon-222 release from 500 years.

stabilized tailings piles per RRY would be 100 Ci in 100 years and 4090 Ci in 500 years and 53,800 Ci in 1000 years.4 The total-body, bone, and bronchial epithelium dose commitments for these periods are as follows:

Population dose commitments (person-rems)

Time span (years)

Releases (Ci)

Total Bone Lung (bronchial body epithelium) 100 100 2.6 68 56 500 4,090 110 2,800 2,300 1,000 53,800 1,400 37,000 30,000 I-6

If risk estimators of 135, 6.9, and 22 cancer deaths per million person-rems for total-budy, bone, and lung exposures, respectively, are used, the estimated risk of cancer mortality resulting from mining, milling, and active tailings emissions of radon-222 is about 0.11 cancer fatalities per RRY. When this risk from radon-222 emissions from stabilized tailings over a 100 year release period is added,'the estimated risk of cancer mortality over a 100 year period is un-changed.

Similarly, a risk of about 1.2 cancer fatalities is estimated over a 1000 year release period per RRY. When potential radon releases from reclaimed and unreclaimed open pit mines are included, the overall risks of radon induced cancer fatal; ues per RRY range as follows:

0.11 to 0.19 fatalities for a 100-year period, 0.19 to 0.57 fatalities for a 500 year period, and 1.2 to 2.0 fatalities for a 1000 year period.

To illustrate:

A single-model 1000-KWe LWR operating at an 80% capacity factor for 30 years would be predicted to induce between 3.3 and 5.7 cancer fatalities in 100 yr, 5.7 and 17 in 500 yr, and 36 and 60 in 1000 yr as a result of releases of radon-222.

These doses and predicted health effects have been compared with those that can be expected from natural-background emissions of radon-222.

Calculated using data from the National Council on Radiation Protection (NCRP)5 the average radon-222 concentration in air in the contiguous United States is about 3

150 pCi/m, which the NCRP estimates will result in an annual dose to the bronchial epithelium of 450 millirems.

For a stabilized future U.S. population of 300 million, this represents a total lung dose commitment of 135 million person-rems per year.

If the same risk estimator of 22.2 lung cancer fatalities per million person-lung rems used to predict cancer fatalities for the model 1000 Mde LWR is used, estimated lung cancer fatalities alone from background radon-222 in the air can be calculated to be about 3000 per year, or 300,000 to 3,000,000 lung cancer deaths over periods of 100 to 1000 years respectively.

In addition to the radon-related potential health effects from the fuel cycle, other nuclides produced in the cycle, such as carbon-14, will contribute to population exposures.

It is estimated that 0.08 to 0.12 additional cancer deaths may occur per RRY (assuming that no cure or prevention of cancer is ever developed) over the next 100 to 1000 years, respectively, from exposures to these other riuclides.

The latter exposures can also be compared with those from naturally occurring terrestrial and cosmic-ray sources.

These average about 100 millirems.

Therefore, for a stable future population of 300 million persons, the whole-body dose commitment would be about 30 million person rems per year, or 3 billion person-rems and 30 billion person-rems for periods of 100 and 1000 years respectively.

These dose commitments could produce aL9ut 400,000 and 4,000,000 cancer deaths during the same time periods.

From the above analysis, the NRC staff concludes that both the dose commitments and health effects of the uranium fuel cycle are insignificant when compared to dose commitments and l

potential health effects to the U.S. population resulting from all l

natural-background sources.

I-7

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• s 6.

Radioactive Wastes The quantities of buried radioactive waste material (low-level, high-level, and transuranic wastes) are specified in Table S-3.

For low-level waste disposal at land burial facilities, the Commission notes in Table S-3 that there will be no significant radioactive releases to the environment.

The Commission notes that high-level and transuranic wastes are to be buried at a Federal Repository and that no release to the environment is associated with such disposal.

NUREG-0116,6 which provides background and context for the high-level and transuranic Table S-3 values established by the Commission, indicates that these high-level and transuranic wastes will be buried and will not be released to the biosphere.

No radiological environmental impact is anticipated from such disposal.

7.

Occupational Dose The annual occupational dose attributable to all phases of the fuel cycle for the model 1000-MWe LWR is about 200 person rems.

The NRC staff concludes that this occupational dose will not have a significant environmental impact.

8.

Transportation The transportation dose to workers and the public is specified in Table S-3.

This dose is small and not considered significant in comparison to the natural-background dose.

9.

Fuel Cycle The staff's analysis of the uranium fuel cycle did not depend on the selected fuel cycle (no recycle or uranium-only recycle), because the data provided in Table S-3 include maximum recycle option impact for each element of the fuel cycle.

Thus the staff's conclusions as to acceptability of the environmental impacts of the fuel cycle are not affected by the specific fuel cycle selected.

REFERENCES FOR APPENDIX I (1) Council on Environmental Quality, "The Seventh Annual Report of the Council on Environmental Quality," September 1976, Figs. 11-27 and 11-28, pp. 238-239.

(2)

U.S. Nuclear Regulatory Commission, " Final Generic Environmental Statement l

on the Use of Recycle Plutonium in Mixed 0xide Fuel in Light-Water-Cooled j

Reactors, Report NUREG-0002, Washington, D.C., August 1976.

(3)

U.S. Department of Energy, " Statistical Data of the Uranium Industry,"

Report GJ0-100(8-78), January 1, 1978.

(4)

R. Gotchy, U.S. Nuclear Regulatory Commission, transcript of direct testimony given "In the Matter of Duke Power Company" (Perkins Nuclear Station), Docket No. 50-448, April 17, 1978.

I I-8 l

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• s 1

(5) National Council on Radiation Protection and Measurements " Natural Background Radiation in the United States," Publication No. 45, November 1975.

(6)

U.S._ Nuclear Regulatory Commission, " Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle," Report NUREG-0116 (Supplement I to WASH-1248), Washington, D.C., October 1976.

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e e

o APPENDIX J EXAMPLES OF SITE-SPECIFIC DOSE ASSESSMENT CALCULATIONS J-1

's 1.

Calculational Approach As mentioned in the text the quantities of radioactive material that may be released annually from the Waterford 3 are estimated on the basis of the description of the radwaste systems in the applicant's ER and FSAR and by using the calculational model and parameters described in NUREG-0017.2 These estimated effluent release values along with the applicant's site and environmental data in the ER and in subsequent answers to NRC staff questions are used in the calculation of radiation doses and dose commitments.

The models and considerations for environmental pathways that lead to estimates of radiation doses and dose commitments to individual members of the public near the plant and of cumulative doses and dose commitments to the entire population within an 80-km radius of the plant as a result of plant operations are discussed in detail in Regulatory Guide 1.109.2 Use of these models with additional assumptions for environmental pathways that lead to exposure to the general population outside the 80-km (50 mile) radius are described in Appendix H of this Statement.

The calculations performed by the staff for the potentially contaminated atmosphere and hydrosphere provide total integrated dose commitments to the entire population within 80 km of the station based on the projected population distribution in the year 2000. The dose commitments represent the total dose that would be received over a 50 yr period, following the intake of radioactivity for 1 yr under the conditions existing 15 years after the station begins operation (i.e., the mid point of station operation).

For younger persons, changes in organ mass and metabolic parameters with age after the initial intake of radioactivity are accounted for.

2.

Dose Commitments from Radioactive Effluent Releases Radioactive effluents released to the atmosphere and to the hydrosphere from the Station will result in very small radiation dose commitments to individual members of the public and to the general population.

The NRC staff estimates of the expected gaseous and particulate releases (listed in Table J-1) and the expected liquid releases (listed in Table J-8) along with the site meteorological and hydrological considerations (summarized in Tables J-2 and J-9 respectively) were used to estimate radiation doses and dose commitments.

Four years of meteorological data were used in the calculation of relative concentrations of effluents.

The data were collected onsite from July 1972 to l

June 1975 and from February 1977 to February 1978.

The long-term diffusion t

estimates were made using the procedure described in Regulatory Guide 1.111, Revision 1.3 Open terrain recirculation factors were used by the staff in the computer model.

(a) Radiation Dose Commitments to Individual Members of the Public As explained in tne text, calculations are made for a hypothetical individual member of the public (i.e., the maximally exposed individual) who would be expected to receive the highest radiation dose from all appropriate pathways.

This method tends to overestimate the doses since assumptions are made that would be difficult for a real individual to fulfill.

J-2

i Individual receptor locations and pathway locations considered for the maximally exposed individual are listed ir. Table J-3.

The estimated dose commitments to the individual who is subject to maximum exposure at selected offsite locations from airborne releases of radioiodine and particulates, and waterborn releases are listed in Tables J-4, J-5, and J-6.

The maximum annual beta and gamma air dose and the maximum total body and skin dose to an individual, at the site boundary, also are presented in Tables J-4, J-5, and J-6.

The maximally exposed individual is assumed to consume well above average quantities of the potentially affected foods and to spend more time at potentially affected locations than the average person as indicated in Tables E-4 and E-5 of Regulatory Guide 1.109.2 With regard to the doses calculated from the nearest farm (ESE 0.6m;) the staff assumed that 20% of the maximum individual's vegetable consumption is obtained from this location.

(b) Cumulative Dose Commitments to the General Population Annual radiation dose commitments from airborne and waterborne radioactive releases from Waterford 3 are estimated for two populations in the year 2000:

(1) all members of the general public within 80 km (50 miles) of the station (Table J-5) and (2) the entire U.S. population (Table J-7).

Dose commitments beyond 80 km are based on the assumptions discussed in Appendix H.

For perspective, annual background radiation doses are given in the tables for both populations.

REFERENCES FOR APPENDIX J 1.

U.S. Nuclear Regulatory Commission, " Calculations of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code) NUREG-0017, U.S. Nuclear Regulatory Commission, April 1976.

2.

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Reg. Guide 1.109, Rev. 1, U.S. Nuclear Regulatory Commission, October 1977.

3.

" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Reactors." Reg. Guide 1.111, Rev. 1, U.S. Nuclear Regulatory Commission, July 1977.

f J-3

• s-Table J-1 Calculated Releases of Radioactive Materials in Gaseous Effluents in Curies per year from Waterford 3 i

Nuclide Plant Stack Plant Stack Turbine Bldg (continuous)

(intermittent)

(continuous)

Kr-85m 5.0 2.0 a

Kr-85 330 73 a

Kr-87 2.0 a

Kr-88 8.0 2.0 a

Xe-131m 8.0 52 a

Xe-133m 10 42 a

Xe-133 730 6400 a

Xe-135 15 12 a

Xe-138 1.0 a

a Total Noble Gases 7692 Mn-54 0.0047 0.000023 b

Fe-59 0.0016 0.0000079 b

Co-58 0.016 0.000079 b

C0-60 0.0073 0.000036 b

Sr-89 0.0034 0.0000018 b

Sr-90 0.00006 0.00000032 b

Cs-134 0.0047 0.000023 Cs-137 0.0078 0.000049 b

Total Particulates 0.04 I-131 0.013 0.0027 0.0041 1-133 0.016 0.00096 0.0035 H-3 940 a

a C-14 7

1 a

a = less than 1.0 Ci/yr for noble gases and carbon-14 less than 10 4 Ci/yr for iodine b = less than 1% of total for this nuclide l

f i

J-4

a Table J-2 Summary of Atmospheric Dispersion Factors (x/Q) and Relative Deposition Values for Maximum Site Boundary and Receptor Locations Near Waterford 3 Relative Location X/Q (sec/m )

Deposition (m 2) 3 Site boundary (ESE 0.6 mi) 1.4 x 10 5 2.3 x 10.s Nearest ** residence and milk cow 7.9 x 10 8 2.3 x 10.s (NW 0.9 mi)

Nearest farm (ESE 0.31 mi) 4.5 x 10 5 6.5 x 10.s Nearest meat animal (NW 0.8 mi) 1.1 x 10 8 3.2 x 10.s A

The values presented in this table are corrected for radioactive decay and cloud depletion from deposition, where appropriate, in accordance with Regulatory Guide 1.111, Rev.1, " Methods for Estimating Atomospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light Water Reactors," July 1977,a nn

" Nearest" refers to that type of location where the highest radiation dose is expected to occur from all appropriate pathways.

Table J-3 Nearest Pathway Locations Used for Maximum Individual Dose Commitments for Waterford 3 Location Sector Distance (mi)

Site boundary

• ESE

0. 6 Residence **

NW 0.9 Farm **

ESE 0.31 Milk cow NW 0.9 Meat animal NW 0.8

• Beta and gamma air doses, total body doses, and skin doses from 9ble gases are determined at site boundaries.

• • Dose pathways including inhalation of atmospheric radioactivity, exposure to deposited radionuclides, and submersion in gaseous radioactivity are evaluated at residences.

J-5

Table J-4 Annual Dose Commitments to a Maximally Exposed Individual Near Waterford 3 Lccation Pathway Doses (mrem /yr per unit)

Noble Gases in Gaseous Effluents Total Body Skin Gamma Air Dose Beta Air Dose (mrad /yr per (mrad /yr per unit) unit)

Near:stgite Direct radiation boundary from plume 1.6 4.6

2. 6 8.1 (ESE 0.6 km)

D Iodine and Particulates in Gaseous Effluents Total Body Organ e

N2arest site Ground deposit 0.29 (T) 0.29 (C) (bone) boundary Inhalation 0.04 (T) 0.006 (C) (bone)

(ESE 0.6 mi)

Nearest farm 2.4 (C) 8.2 (C) (bone)

(ESE 0.31 mi) Vegetable consumption N arest milk Ground deposit 0.21 (C) 0.21 (C) (bone) cow Inhalation 0.27 (C) 0.004 (C) (bone)

(NW 0.9 mi)

Vegetable consumption 2.10 (C) 7.4 (C) (bone)

Cow milk consumption 0.96 (C) 3.6 (C) (bone)

Meat consumption 0.25 (C) 1.1 (C) (bone)

Nearest meat Meat consumption 0.40 (C) 1.7 (C) (bone) cnical (NW 0.8 mi)

Liquid Effluents (Adults)

Nearest Drinking Water Ingestion Total Body Organ Water

<0.01 0.03 (thyroid)

(St. Charles Parish)

Nearest fish fish ingestion 0.08 0.11 (liver)

(Discharge)

"" Nearest" refers to that site boundary location where the highest radiation doses as a result of gaseous effluents have been estimated to occur.

Dos 2s are for the age group that results in the highest dose:

T= teen, C= child, I= infant.

c" Nearest" refers to the location where the highest radiation dose to an t

individual from all applicable pathways has been estimatcd.

l J-6

O, '. s.O Table J-5 Calculated Appendix I Dose Commitments to a Maximally Exposed Individual and to the Population frcm Operation of Waterford 3 Annual Dose per Reactor Unit Individual Appendix I Calculated a

Design Objectives Doses Liquid effluents Dose to total body from all pathways 3 mrem 0.1 mrem Dose to any organ from all pathways 10 mrem 0.12 mrem Noble gas effluents (at site boundary)

Gamma dose in air 10 mrad 2.6 mrad Beta dose in air 20 mrad 8.1 mrad Dose to total body of an individual 5 mrem 1.6 mrem Dose to skin of an individual 15 mrem 4.6 mrem b

Radiciodines and particulates Dose to any organ from all pathways 15 mrem 12 mrem (bone-child)

Population Within 80 km i

Total Body Organ (person-rem) c Natural-background radiation 180,000 Liquid effluents 6.0 7.1 (thyroid)

Noble gas effluents 0.23 0.23 (bone)

Radioiodine and particulates 5.5 8.7 (bone) a0esign Objectives from Sections II.A, II.8, II.C, and II.D of Appendix I, 10 CFR Part 50 consider doses to maximum individual and population per reactor unit.

bCarbon-14 and tritium have been added to this category.

c" Natural Radiation Exposure in the United States," U.S. Environmental i

Protection Agency, ORP-SID-72-1, June 1972; using the average back-ground dose for Louisiana of 84 mrem /yr, and year-2000 projected population of 2,182,000.

l J-7

Table J-6 CalculatedRM-50-2DoseCommigmentstoaMaximallyExposedIndividual from Operation of Waterford 3

~

Annual Dose per Site RM-50-2 Calculated b

Design Objectives Doses Liquid effluents Dose to total body or any organ from all pathways 5 mrem 0.1 mrem Activity-release estimate, excluding tritium (Ci/ unit) 5 0.24 Noble gas effluents (at site boundary)

Gamma dose in air 10 mrad 2.6 mrad Beta dose in air 20 mrad 8.1 mrad Dose to total body of an individual 5 mrem 1.6 mrem c

Radioiodin'e and particulates Dose to any organ from all pathways 15 mrem 12 mrem (bone)

I-131 activity release (Ci) 1 0.4

1. n optional method of demonstrating compliance with the cost-t.enefit Section (II.D)

A of Appendix I to 10 CFR Part 50.

bAnnex to Appendix I to 10 CFR Part 50.

CCarbon-14 and tritium have been added to this category.

l J-8

• *,. o Table J-7 Annual Total-Body Population Dose Commitments, Year 2000 U.S. Population Category Dose Commitment, person rem /yr a

a Natural background radiation 27,000,000 Waterford Nuclear Station Unit 3 operation Plant workers 440 General public:

b Liquid effluents 11.

Gaseous effluents 42 Transportation of fuel and waste 7

ausing the average U.S. background dose (100 mrem /yr) and year 2000 projected U.S. p"opulation from " Population Estimates and Projections, Series II, U.S. Department of Commerce, Bureau of the Census, Series P-25, No. 541 February 1975.

b80-km (50-mile) population dose l

l i

J-9

c'..,. e Table J-8 Calculated Release of Radioactive Materials in Liquid Effluents.from Waterford 3 Nuclide Ci/yr NucTide Ci/yr Corrosion & Activation Products Cr-51 0.00007 I-130 0.00021 Mn-54 0.001 Te-131m 0.00005 Fe-55 0.00006 I-131 0.092 Fe-59 0.00004 Te-132 0.00072 Co-58 0.0046 I-132 0.0042 Co-60 0.0088 I-133 0.058 Zr-95 0.0014 I-134 0.00002 Nb-95 0.002 Cs-134 0.015 Np-239 0.003 I-135 0.0096 Te-129 0.00003 Fission Products B r-83 0.00004 Cs-136 0.0007 Sr-89 0.00001 Cs-173 0.026 Mo-99 0.0024 Bs-137m 0.0015 Tc-99m 0.0028 Ce-144 0.0052 a

Ru-103 0.00014 All others 0.00006 Ru-106 0.0024 Total except Ag-110m 0.00044 tritium 0.24 Te-127 0.00002 Te-129m 0.00005 Tritium release aNuclides whose release rates are less than 10 5 Ci/yr are not listed individually but are included in the category "All others."

J-10

h.

Table J-9 Summary'of Hydrologic Transport and Dispersion for Liquid Releases from Waterford 3 i

Transit Time Dilution Location (hours)

Factor Nearest drinking

1. 0 5

water intake (Union Cartide)

(-2.6 mi, downstream)

Nearest sport fishing location (plant discharge) 0.01 1

Nearest shoreline 0.01 1

(plant discharge)

Nearest irrigated crops (St. Charles) 0.1 5

aSee Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

J-11

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