Regulatory Guide 3.35

Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Plutonium Processing and Fuel Fabrication Plant

ML12220A062
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Issue date:	07/31/1979
From:	Office of Nuclear Regulatory Research, NRC/OSD
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RG-3.035, Rev. 1
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Revision 1 U.S. NUCLEAR REGULATORY COMMISSION July 1979

• REGULATORY GUIDE

OFFICE OF STANDARDS DEVELOPMENT

REGULATORY GUIDE 335 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL

CONSEQUENCES OF ACCIDENTAL NUCLEAR CRITICALITY IN A

PLUTONIUM PROCESSING AND FUEL FABRICATION PLANT

A. INTRODUCTION

will review the proposal and approve its use, if found acceptable.

Section 70.22, "Contents of Applications," of

10 CFR Part 70, "Domestic Licensing of Special

B. DISCUSSION

Nuclear Materials," requires, that each appli- cation for a license to possess and use special In the process of reviewing applications for nuclear material in a plutonium processing and permits and licenses authorizing the construc- fuel fabrication plant contain a description and tion or operation of plutonium processing and safety assessment of the design bases of the fuel fabrication plants, the NRC staff has principal structures, systems, and components developed a number of appropriately conser- of the plant. Section 70.23(a)(3) states that vative assumptions that are used by the staff applications will be approved if the Commission to evaluate an estimate of the radiological determines that, among other factors, the consequences of various postulated accidents.

applicant's proposed equipment and facilities These assumptions are based on previous are adequate to protect health and minimize accident experience, engineering judgment, danger to life and property, and Sec- and on the analysis of applicable experimental tion 70.23(b) states that the Commission will results from safety research programs. This approve construction of the principal struc- guide lists assumptions used by the staff to tures, systems, and components of the plant evaluate the magnitude and radiological conse- when the Commission has determined that the quences of a criticality accident in a plutonium design bases of the principal structures, sys- processing and fuel fabrication plant.

tems, and components and the quality assurance program provide reasonable A criticality accident is an accident resulting assurance of protection against the in the uncontrolled release of energy from an consequences of potential accidents. assemblage of fissile material. The cir- cumstances of a criticality accident are difficult In plutonium processing and fuel fabrication to predict. However, the most serious plants, a criticality accident is one of the criticality accident would be expected to occur postulated accidents used to evaluate the ade- when the reactivity (the extent of the deviation quacy of an applicant's proposed activities with from criticality of a nuclear chain reacting respect to the public health and safety. This medium) could increase most rapidly and guide describes methods used by the NRC staff without control in the fissile accumulation of in the analysis of such accidents. These the largest credible mass. In plutonium pro- methods result from review and action on a cessing and fuel fabrication plants where con- number of specific cases and, as such, reflect ditions that might lead to criticality are the lates~t general NRC-approved approaches to carefully avoided because of the potential for the problem. If an applicant desires to employ adverse physical and radiological effects, such new information that may be developed in the an accident is extremely uncommon. However, future or to use an alternative method, NRC experience with these and related facilities has demonstrated that criticality accidents may

• Lines indicate substantive changes from previous issue. occur.

USNRC REGULATORY GUIDES Comments should be sent to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, D.C. 20566, Attention: Docketing and Regulatory Guides are issued to describe and make available to the public Service Branch.

methods acceptable to the NRC staff of implementing specific parts of the Commission's regulations, to delineate techniques used by the staff in evalu- The guides are issued in the following ten broad divisions:

sting specific problems or postulated accidents, or to provide guidance to applicants. Regulatory Guides are not substitutes for regulations, and com- 1. Power Reactors 6. Products phiance with them is not required. Methods and solutions different from tdose 2. Research and Test Reactors 7. Transportation set out in the guides will be acceptable if they provide a basis for the findings 3. Fuels and Materials Facilities 8. Occupational Health requisite to the issuance or continuance of a permit or license by the 4. Environmental and Siting 9. Antitrust and Financial Review Commission. 5. Materials and Plant Protection 10. General Requests for single copies of issued guides (which may be reproduced) or for Comments and suggestions for improvements in thease guides are encouraged at placement on an automatic distribution list for single copies of future guides all times, and guides will be revised, as appropriate, to accommodate comments in specific divsions should be made in writing to the U.S. Nuclear Regulatory and to reflect new information or experience. This guide was revised as a result Commission, Washington, D.C. 20555, Attention: Director, Division of of substantive comments received from the public and additional staff review. Technical Information and Document Control.

In plutonium processing and fuel fabrication and dense layers, loss of water moderator by plants, such an accident might be initiated by boiling, or expulsion of part of the mass.

(1) the inadvertent transfer or leakage of a solution of fissile material from a geometrically Generally, the criticality incidents were safe containing vessel into an area or vessel characterized by an initial burst or spike in not so designed, (2) introduction of excess the curve of fission rate versus time followed fissile material solution to a vessel, (3) intro- by a rapid but incomplete decay of the fission duction of excess fissile material to a solution, rate as the shutoff mechanism was initiated. As

(4) overconcentration of a solution, (5) preci- more than one shutdown mechanism may affect pitation of fissile solids from a solution and the reactivity of the system and the effect of a their retention in a vessel, (6) introduction of particular mechanism may be counteracted, the neutron moderators or reflectors (e.g., initial burst was frequently succeeded by a entrance of water to a higly under-moderated plateau period of varying length. This plateau system), (7) deformation of or failure to was characterized by a lesser and declining maintain safe storage arrays, or (8) similar fission rate and finally by a further dropoff as actions which can lead to increases in the shutdown was completed. The magnitude of the reactivity of fissile systems. Some acceptable initial burst was directly related to the rate of means for minimizing the likelihood of such increase of reactivity and its magnitude above accidents are described in Regulatory the just-critical value but was inversely related Guide 3.4, "Nuclear Criticality Safety in to the background neutron flux, which is much Operations with Fissionable Materials Outside greater for plutonium than for uranium Reactors. "1 systems.

I. CRITICALITY ACCIDENT EXPERIENCE IN RELATION TO

THE ESTIMATION OF THE MOST SEVERE ACCIDENT

Those systems consisting only of solid fissile, reflector, or moderator materials exhibited little or no plateau period, whereas Stratton (Ref. 1) has reviewed in detail

34 occasions prior to 1966 when the power level solution systems had well developed plateaus.

For solution systems, the energy release of a fissile system increased without control as a result of unplanned or unexpected changes in during the plateau period, because of its dura- tion, provided the major portion of total energy its reactivity. Although only six of these released. For purposes of the planning neces- incidents occurred in processing operations, sary to deal adequately with criticality and the remainder occurred mostly in facilities for obtaining criticality data or in experimental incidents in experimental and production-type nuclear facilities, Woodcock (Ref. 3) made use reactors, the information obtained and its of these data to estimate possible fission yields correlation with the characteristics of each from excursions in various types of systems.

system have been of considerable value for use For example, spike yields of 1E+17 and 1E+18 in estimating the consequences of accidental and total yields of 3E+18 and 3E+19 fissions criticality in process systems. The incidents were suggested for criticality accidents occurred in aqueous solutions of uranium or occurring in solution systems of 100 gallons or plutonium (10), in metallic uranium or less and more than 100 gallons, respectively.

plutonium in air (9), in inhomogeneous water- Little or no mechanical damage was predicted at moderated systems (9), and in miscellaneous these levels.

solid uranium systems (6). Five occurred in plutonium systems, including reactors and

2. METHODS DEVELOPED FOR PREDICTING THE MAGNITUDE

criticality studies, of which three were in OF CRITICALITY ACCIDENTS

solutions.

The nuclear excursion behavior of solu- The estimated total number of fissions per tions of enriched uranium has been studied incident ranged from 1E+15 2 to 1E+20 with a extensively both theoretically and experi- median of about 2E+17. More recently, another mentally. A summary by Dunenfeld and Stitt incident in a plutonium processing facility at (Ref. 4) of the kinetic experiments on water Windscale (U.K.) was described in which a -J

boilers, using uranyl sulfate solutions, total yield of about 1E+15 fissions apparently describes the development of a kinetic model occurred (Ref. 2). In ten cases, the that was confirmed by experiment. This model supercriticality was halted by an automatic defines the effects of thermal expansion and control device. In the remainder, the shutdown radiolytic gas formation as power-limiting and was effected as a consequence of the fission shutdown mechanisms.

energy release which resulted in thermal expansion, density reduction from the The results of a series of criticality excur- formation of very small bubbles, mixing of light sion experiments resulting from the introduc- tion of uranyl nitrate solutions to vertical

'Copies may be obtained from the U.S. Nuclear Regulatory cylindrical tanks at varying rates are sum- Commission, Washington, D.C. 20555, Attention; Director, marized by Ldcorchd and Seale (Ref. 5). This Division of Document Control. report confirms the applicability of the kinetics

2

1E÷15 = 1 x 1015. This notational form will be used in this model for solutions, provides correlations of guide. peak power with reactivity addition rate, notes

3.35-2

the importance of a strong neutron source in 3E+22 fissions resulting in a serious explosion limiting peak power, and indicates the nature could be conceived for large storage arrays of the plateau following the peak. where prompt criticality was exceeded, e.g.,

by collapse of shelving. It is recognized that in Many operations with fissile materials in a such arrays, where reactivity is more likely to plutonium processing plant may be conducted be increased by the successive additions of with aqueous (or organic solvent) solutions of small increments of materials, only a delayed fissile materials. Consequently, well-founded critical condition with maximum yields of 1E+19 methods for the prediction of total fissions and fissions is likely. These estimates could aid in maximum fission rate for accidents that might the analysis of situations in plant systems.

occur in solutions (in process or other vessels) However, they should not be taken as absolute by the addition of fissile materials should be of values for criticality assumptions for the considerable value in evaluating the effects of purpose of this guide.

possible plutonium processing plant criticality accidents. From the results of excursion For systems other than solution systems, studies and from accident data, Tuck (Ref. 6) the estimation of the peak fission rate and the has developed methods for estimating (1) the total number of fissions accompanying an acci- maximum number of fissions in a 5-second dental nuclear criticality may be estimated with interval (the first spike), (2) the total number the aid of information derived from accident of fissions, and (3) the maximum specific fis- experience and from the SPERT-l reactor tran- sion rate in vertical cylindrical vessels, 28 to sient tests with light- and heavy-water

152 cm in diameter and separated by >30 cm moderated uranium-alumium and U0 2 -stainless from a bottom reflecting surface, resulting steel clad fuels (Ref. 8). Oxide core tests in from the addition of up to 500 g/1l solutions of the latter group provide some information on Pu-239 or U-235 to the vessel at rates of 0.7 to energy release mechanisms that may be

7.5 gal/min. Tuck also gives a method for effective, for example, in fabricated fuel estimating the power level from which the element storage in a mixed oxide fuel fabrica- steam-generated pressure may be calculated tion plant. Review of unusal process struc- and indicates that use of the formulas for tanks tures, systems, and components for the

>152 cm in diameter is possible with a loss in possibility of. accidental criticality should also accuracy. consider recognized anomalous situations in which the possibility of accidental nuclear cri- Methods for estimating the number of fis- ticality may be conceived (Ref. 9).

sions in the initial burst and the total number of fissions, derived from the work reported by The application of the double-contingency L6corchi and Seale (Ref. 5), have also been principle3 to fissile material processing opera- developed by Olsen and others (Ref. 7). These tions has been successful in reducing the were evaluated by application to ten actual probability of accidental criticality to a low accidents that have occurred in solutions and value. As a consequence, the scenarios were shown to give conservative estimates in required to arrive at accidental criticality all cases except one. involve the assumption of multiple breakdowns in the nuclear criticality safety controls. It has Fission yields for criticality accidents therefore been a practice to simply and occurring in solutions and some heterogeneous conservatively as'sume an accidental criticality systems, e.g., aqueous/fixed geometry, can be of a magnitude equal to, or some multiple of, estimated with reasonable accuracy using the historical maximum for all criticality acci- existing methods. However, methods for esti- dents outside reactors without using any mating possible fission yield from .other types scenario clearly defined by the specific opera- of heterogeneous systems, e.g., aqueous/ tions being evaluated. In the absence of powder, are less reliable because of the sufficient guidance, there has been wide vari- uncertainties involved in predicting the ation in the credibility of the postulated reactivity rate. The uncertainty of geometry magnitude of the occurrence (particularly the and moderation results in a broad range of size of the initial burst), the amount of energy possible yields. and radioactivity assumed to be released, and the magnitude of the calculated consequences.

Woodcock (Ref. 3) estimated that in solid plutonium systems, solid uranium systems, and It is the staff's judgment that the evalua- heterogeneous liquid/powder systems (fissile tion of the criticality accident should assume material not specified) total fission yields (sub- the simultaneous breakdown of at least two stantially occurring within the spike) of 1E+18, independent controls throughout all elements of

3E+19, and 3E+20, respectively, could be the operation. Each control should be such that predicted. Mechanical damage varied from its circumvention is of very low probability.

slight to extensive. Heterogeneous systems Experience has shown that the simultaneous consisting of metals or solids in water were

3The double-contingency principle is defined in ANSI N16. 1- estimated to achieve a possible magnitude of

1975, "Nuclear Criticality Safety in Operations with Fissionable

1E+19 following an initial burst of Materials Outside Reactors," which is endorsed by Regulatory

3E+18 fissions. The possibility of a burst of Guide 3.4.

3.35-3

failure of two independent controls is very the adequacy of structures, systems, and unlikely if the controls are derived, applied, components provided for the prevention or and maintained with a high level of quality mitigation of the consequences of accidents, assurance. However, if controls highly the applicant should evaluate credible dependent on human actions are involved, this criticality accidents in all those elements of the approach will call for some variation in the plant provided for the storage, handling, or assumed number of control failures. The processing of fissile materials or into which criticality accidents so conceived should then fissile materials in significant amounts could be be analyzed to determine the most severe introduced. To determine the circumstances of within the framework of assumed control the criticality accidents, controls judged failures, using realistic values of such equivalent to at least two highly reliable, variables as the fissile inventory, vessel sizes, independent criticality controls should be and pump transfer rates. assumed to be circumvented. The magnitude of the possible accidents should then be assessed,

3. RADIOLOGICAL CONSEQUENCES OF ACCIDENTAL CRITI- on an individual case basis, to estimate the CALITY extent and nature of possible effects and to provide source terms for dose calculations. The Past practice has been to evaluate the most severe accident should then be selected radiological consequences to individuals of for the assessment of the adequacy of the postulated accidental criticality in plutonium plant. In order to determine the source terms processing and fuel fabrication plants in terms for release of plutonium, the powder mixture of a fraction of the guideline values in 10 CFR should be the maximum weight percent pluto- Part 100, "Reactor Site Criteria." nium to uranium compound to be used in a mixed oxide fuel fabrication plant.

The consequences of a criticality accident may be limited by containment, shielding, Calculation of the radioactivity of fis- isolation distance, or evacuation of adjacent sion products may be accomplished by computer occupied areas subsequent to detection of the code RIBD (Ref. 10). An equivalent calculation accident. If the impact of a criticality accident may be substituted, if justified on an is to be limited through evacuation of adjacent individual case basis.

occupied areas, there should be prior formal arrangements with individual occupants and b. If the results of the preceding evalu- local authorities sufficient to ensure that such ation indicate that no possible criticality movements can be effected in the time allowed. accident exceeds in severity the criticality accident postulated in this section, then the The equations provided for estimating conditions of the following example may be doses from prompt gamma and neutron radiation assumed for the purpose of assessing the were developed using experimental and adequacy of the facility. A less conservative historical data. The report, "Promp Neutron set of conditions may be used if they are shown and Gamma Doses from an Accidental to be applicable by the specific analyses Criticality," explains this development.* These conducted in accordance with paragraph C.l.a equations cannot be expected to be as accurate above.

as detailed calculations based on actual accident conditions. Comparisons with An excursion that produces an initial published information indicate they may not be burst of 1E+18 fissions in 0.5 seconds followed conservative for smaller accidents .(e.g. , 1- successively at 10-minute intervals by

2E+17 fissions). However, for accidents that 47 bursts of 1.9E+17 fissions for a total of are likely to be assumed for safety assessment 1E+19 fissions in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is assumed to occur.

purposes, they appear to be sufficiently The excursion is assumed to be terminated by conservative. These equations are included in evaporation of 100 liters of the solution.

the guide to provide a simplified method for estimatinK prompt gamma and neutron radiation 2. ASSUMPTIONS RELATED TO THE RELEASE OF RADIO-

doses from a potential criticality accident. ACTIVE MATERIAL ARE AS FOLLOWS: 4

C. REGULATORY POSITION

a. It should be assumed that all of the noble gas fission products and 25% of the iodine I. FOLLOWING ARE THE PLANT ASSESSMENT AND ASSUMP- radionuclides are released directly to a TIONS RELATED TO ENERGY RELEASE FROM A CRITI- ventilated room whose construction is typical of CALITY ACCIDENT AND THE MINIMUM CRITICALITY the plant's Class I structures. If the accident ACCIDENT TO BE CONSIDERED: is assumed to occur in a solution, it should also be assumed that an aerosol, which is generated a. When defining the characteristics of an from the evaporation of solution during the assumed criticality accident in order to assess excursion, is released directly to the room atmosphere. The aerosol should be assumed to

• A copy of Charles A. Willis' report, "Prompt Neutron and Gamma Doses, from an Accidental Criticality," is available for 4Certain assumptions for release of radioactive material, dose inspection at the NRC Public Document Room, 1717 H Street conversion, and atmospheric diffusion reflect the staff's NW., Washington, D.C. position indicated in Regulatory Guide 1.3 (Ref. 20).

3.35-4

comprise 0.05% of the salt content of the first foot, and a factor of 5.5 for each addi- solution that is evaporated. The room volume tional foot.

and ventilation rate and retention time should be considered on an individual case basis. (2) Prompt Neutron Dose

2 b. The effects of radiological decay during Dn = 7E-20 Nd" e-5.2d transit within the plant and in the plant where exhaust system should be taken into account on an individual case basis.

Dn = neutron dose (rem)

c. The reduction in the amount of radio- N = number of fissions active material available for release to the environment through the plant stack as a d = distance from source (kin)

result of the normal operation of filtration systems in the plant exhaust systems may be For concrete, the dose should be taken into account, but the amount of reduc- reduced by a factor of 2.3 for the first 8 tion in the concentration of radioactive mate- inches, 4.6 for the first foot, and a factor of rials should be evaluated on an individual case

20 for each additional foot.

basis.

d. Table 1 lists the radioactivity of sig- b. No correction should be made for deple- nificant nuclides released, but it does not tion from the effluent plume of radioactive include the iodine depletion allowance. iodine due to deposition on the ground or for the radiological decay of iodine in transit.

• 3. ACCEPTABLE ASSUMPTIONS FOR DOSE AND DOSE CON-

VERSION ARE AS FOLLOWS:

c. For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing a. The applicant should show that the con- rate of a person off site should be assumed to be 3.47E-4 mS/sec. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> follow- sequences of the prompt gamma and neutron ing the accident, the breathing rate should be dose are sufficiently mitigated to allow assumed to be 1.75E-4 m 3 /sec. These values occupancy of areas necessary to maintain the were developed from the average daily breath- plant in a safe condition following the accident. 3 ing rate (2E + 7 cm /day) assumed in the The applicant should estimate the prompt report of ICRP Committee 11-1959 (Ref. 12).

gamma and neutron doses that could be received at the closest site boundary and nearest residence. The following semi-empirical d. External whole body doses should be equations may be used for these calculations.

calculated using "Infinite Cloud" assumptions, Because detailed evaluations will be dependent i.e., the dimensions of the cloud are assumed on the site and plant design, different methods to be large compared to the distance that the may be substituted on an individual case basis.

gamma rays and beta particles travel. "Such a Potential total dose attenuation due to shielding cloud would be considered an infinite cloud for and dose exposures should be evaluated on an a receptor at the center because any additional individual case basis.

(gamma and] beta emitting material beyond the cloud dimensions would not alter the flux of (I) Prompt 5 Gamma Dose

[gamma rays and] beta particles to the

2 e-3.4d receptor." [See Meteorology and Atomic D = 2.IE-20 Nd- w Energy--1968 (Ref. 13), Section 7.4.1.1;

where editorial additions made so that gamma and beta emitting material could be considered.] Under these conditions, the rate of energy absorption D ¥ = gamma dose (rein)

per unit volume is equal to the rate of energy released per unit volume. For an infinite N = number of fissions uniform cloud containing X curies of beta radioactivity per cubic meter, the beta dose d = distance from source (kin)

rate in air at the cloud center is Data presented in The Effects of Nuclear Weapons (Ref. 11, p. 384) may be used to D- = 0.457EPX

develop dose reduction factors. For concrete, the dose should be reduced by a factor of 2.5 for the first 8 inches, a factor of 5.0 for the The surface body dose rate from beta emitters in the infinite cloud can be approximated as Syost of the gamma radiation is emitted in the actual fission process. Some gamma radiation is produced in various second- being one-half this amount (i.e., pDoo = 0.23EX).

ary nuclear processes, including decay of fission products. For For gamma emitting material, the dose rate in the purposes of this guide, "prompt" gamma doses should be air at the cloud center is evaluated including the effects of decay of significant fission products during the first minute of the excursion. For conditions cited in the example, the equation given includes these considerations. YDo, = o.5o07E X

3.35-5

From a semi-infinite cloud, the gamma dose rate absorbed into the body. For the purpose of in air is this guide, the following assumptions should be made:

= o.25EYx

(1) The radionuclide dose conversion where factors are as recommended by the report of Committee 11, ICRP (Ref. 12) or other appro- I priate source.

D-= beta dose rate from an infinite cloud (rad/sec) (2) The effective half-life for the nu- clide is as recommended in ICRP Publication 6 Do, = gamma dose rate from an infinite (Ref. 16) or other appropriate source.

¥ cloud (rad/sec)

(3) The plutonium and other actinide E = average beta energy per disintegration nuclide clearance half time, or fraction of nu- (MeV/dis) clide clearing the organ, is as recommended by the ICRP task group on lung dynamics EY

¥= average (MeV/dis)gamma energy per disintegration (Ref. 17). A computer code, DACRIN

(Ref. 18), is available for this model. Task group lung model (TGLM) clearance parameters X = concentration of beta or gamma emitting are presented in Table 2; the model is shown isotope in the cloud (Ci/m 3 ) schematically in Figure 2.

e. The following specific assumptions are g. The potential dose exposure for all sig- acceptable with respect to the radioactive cloud nificant nuclides should be estimated for the dose calculations: population distribution on a site-related basis.

(1) The dose at any distance from the 4. ACCEPTABLE ASSUMPTIONS FOR ATMOSPHERIC DIFFU-

plant should be calculated based on the maxi- SION ARE AS FOLLOWS:

mum concentration time integral (in the course of the accident) in the plume at that distance, a. Elevated releases should be considered taking into account specific meteorological, to be at a height equal to not more than the topographical, and other characteristics that actual stack height. 6 Certain site-dependent may affect the maximum plume concentration. conditions may exist, such as surrounding These site-related characteristics should be elevated topography or nearby structures, that evaluated on an individual case basis. In the will have the effect of reducing the actual case of beta radiation, the receptor is assumed stack height. The degree of stack height to be exposed to an infinite cloud at the reduction should be evaluated on an individual maximum ground level concentration at that case basis.

distance from the plant. In the case of gamma radiation, the receptor is assumed to be Also, special meteorological and geo- exposed to only one-half the cloud owing to the graphical conditions may exist that can con- presence of the ground. The maximum cloud tribute to greater ground level concentrations concentration should always be assumed to be in the immediate neighborhood of a stack. For at ground level. example, fumigation should always be assumed to occur; however, the length of time that a

(2) The appropriate average beta and fumigation condition . exists is strongly gamma energies emitted per disintegration may dependent on geographical and seasonal factors be derived from the Table of Isotopes (Ref. 14) and should be evaluated on a case-by-case or other appropriate sources, e.g. , Ref. 23. basis.' (See Figure 3 for elevated releases under fumigation conditions.)

(3) The whole body dose should be considered as the dose from gamma radiation at b. For plants with stacks, the atmospheric I

a depth of 5 cm and the genetic dose at a diffusion model should be as follows:

depth of 1 cm. The skin dose should be the sum of the surface, gamma dose and the beta Scredit for an elevated release should be given only if the point of release is (1) more than two and one-half times the dose at a depth of 7 mg/cm2 . The beta skin height of any structure close enough to affect the dispersion of dose may be estimated by applying an energy- the plume or (2) located far enough from any structure that dependent attenuation factor (Dd/DB) to the could have an effect on the dispersion of the plume. For those surface dose according to a method developed plants without stacks, the atmospheric diffusion factors assuming ground level releases, as shown in Regulatory by Loevinger, Japha, and Brownell (Ref. 15). Position 4.c, should be used.

(See Figure 1.) 7 For sites located more than 2 miles from large bodies of water, such as oceans or one of the Great Lakes, a fumigation f. The "critical organ" dose from the in- condition should be assumed to exist at the time of the accident haled radioactive materials should be estimated. and continue one-half hour. For sites located less than 2 miles from large bodies of water, a fumigation condition should be The "critical organ" is that organ that receives assumed to exist at the time of the accident and continue for the highest radiation dose after the isotope is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3.35-6

(1) The basic equation for atmospheric suming various stack diffusion from an elevated release is heights] windspeed 1 m/ sec;

uniform direction.

exp(-he 2 /2Cz 2

)

X/Q = 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> See Figure 5 for Envelope of- iiua a Pasquill diffusion categories;

yz windspeed 1 m/sec; variable where direction within a 22.50

x = the short-term average centerline value sector.

3 of the ground level concentration (Ci/m ) c. If no onsite meteorological data are available for facilities exhausted wihout stacks, Q = rate of material release (Ci/sec) or with stacks that do not meet the elevated release criteria, the atmospheric diffusion u = windspeed (m/sec) model should be as follows:

a = the horizontal standard deviation of the (1) The 0-to-8 hour ground level re- Y plume (m). [See Ref. 19, Figure V-l, lease concentrations may be reduced by a p. 48.] factor ranging from one to a maximum of three (see Figure 6) for additional dispersion a = the vertical standard deviation of the produced by the turbulent wake of a major z plume (m). [See Ref. .19, Figure V-2, building in calculating nearby potential expo- p. 48.] sures. The volumetric building wake correction factor, as defined in Section 3.3.5.2 of

8 he = effective height of release (m) Meteorology and Atomic Energy--1968 (Ref. 13), should be used in the 0-to-8 hour

(2) For time periods of greater than 8 period only; it is used with a shape factor of hours, the plume from an elevated release one-half and the minimum cross-sectional area should be assumed to meander and spread of a major building only.

uniformly over a 22.50 sector. 9 The resultant equation is (2) The basic equation for atmospheric diffusion from a ground level point source is

2.032 exp(-h e2/2 z2)

1 x/Q = UX

e x/Q= nuraa y ux yz z

where where x = distance from the release point (m); X = the short-term average centerline value other variables are as given in b(l). of the 3 ground level concentration (Ci/m )

0

(3) The atmospheric diffusion model'

for an elevated release as a function of the Q = rate of material release (Ci/sec)

distance from the plant is based on the infor- mation in the following table. u = windspeed (m/sec)

a = the horizontal standard deviation of the Time Following y plume (m) [see Ref. 19, Figure V-i, Accident Atmospheric Conditions p. 48]

0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> See Figure 4 for Envelope of a = the vertical standard deviation of the Pasquill diffusion categories z plume (m) [see Ref. 19, Figure V-2,

[based on Figure A7, p. 481 Meteorology and Atomic Energy--1968 (Ref. 13), as- (3) For time periods of greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the plume should be assumed to

8h =h -h , where h is the height of the release above plant meander and spread uniformly over a 22.50

grads, SanA ht is tde maximum terrain height, above plant sector. 9 The resultant equation is grade, between the point of release and the point at which the calculation is made, he should not be allowed to exceed hs.

2.032 gThe sector may be assumed to shift after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> if local x/Q= a ux meteorological data are available to justify a wind direction z change. This should be considered on an individual case basis.

'l°n some cases, site-dependent parameters such as meteor- where ology, topography, and local geography may dictate the use of a more restrictive model to ensure a conservative estimate of potential offsite exposures. In such cases, appropriate site- X = distance from point of release to the related meteorology should be developed on an indivdual case receptor; other variables are as given basis. in c(2).

3.35-7

(4) The atmospheric diffusion model' 0

D. IMPLEMENTATION

for ground level releases is based on the infor- mation in the following table. The purpose of this section is to provide information to applicants and licensees regard- ing the staff's plans for using this regulatory guide.

Time Following Accident Atmospheric Conditions Except in those cases in which the applicant proposes an alternative method for complying

0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Pasquill Type F, windspeed with specified portions of the Commission's

1 m/sec, uniform direction regulations, the method described herein will be used in the evaluation of submittals for

8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Pasquill Type F, windspeed special nuclear material license applications

1 m/sec, variable direction docketed after December 1, 1977.

within a 22.50 sector.

If an applicant wishes to use this regulatory

• guide in developing submittals for applications

(5) Figures 7A and 7B give the ground docketed on or before December 1, 1977, the level release atmospheric diffusion factors pertinent portions of the application will be based on the parameters given in c(4). evaluated on the basis of this guide.

I

3.35-8

REFERENCES

1. W. R. Stratton, "Review of Criticality 14. C. M. Lederer, J. M. Hollander, I. Perl- Incidents," LA-3611, Los Alamos Scientific man, Table of Isotopes, 6th Edition, Laboratory (Jan. 1967). Lawrence Radiation Laboratory, Univ. of California, Berkeley, California (1967).

2. T. G. Hughes, "Criticality Incident at Windscale," Nuclear Engineering Inter- 15. Radiation Dosimetry, G. J. Hine and G. L.

national, Vol. 17, No. 191, pp. 95-7 Brownell, Editors, Academic Press, New (Feb. 1972). York (1956).

3. E. R. Woodcock, "Potential Magnitude of 16. Recommendations of ICRP, Publication 6, Criticality Accidents," AHSP(RP) R-14, Pergamon Press (1962).

United Kingdom Atomic Energy Authority.

17. "The Metabolism of Compounds of Plutonium

4. M. S. Dunenfeld, R. K. Stitt, "Summary and Other Actinides," a report prepared Review of the Kinetics Experiments on by a Task Group of Committee II, ICRP,

Water Boilers." NAA-SR-7087, Atomic Pergamon Press (May 1972).

International (Feb. 1973).

18. J. R. Houston, D. L. Strenge, and E. C.

5. P. Lgcorch6, R. L. Seale, "A Review of the Watson, "DACRIN--A Computer Program Experiments Performed to Determine the for Calculating Ocean Dose from Acute or Radiological Consequences of a Criticality Chronic Radionuclide Inhalation," BNWL-B-

Accident, " Y-CDC-12, Union Carbide Corp. 389(UC-4), Battelle Memorial Institute, (Nov. 1973). Pacific Northwest Laboratories, Richland, Washington, (Dec. 1974).

6. G. Tuck, "Simplified Methods of Estimating the Results of Accidental Solution Excur- 19. F. A. Gifford, Jr., "Use of Routine sions," Nucl. Technol., Vol. 23, p. 177 Meteorological Observations for Estimating

(1974). Atmospheric Dispersion," Nuclear Safety, Vol. 2, No. 4, p. 48 (June 1961).

7. A. R. Olsen, R. L. Hooper, V. 0. Uotinen, C. L. Brown, "Empirical Model to Estimate Energy Release from Accidental Criticality," 20. Regulatory Guide 1.3, "Assumptions Used for Evaluating the Radiological Consequences ANS Trans., Vol. 19, pp. 189-91 (1974).

of a Loss of Coolant Accident for Boiling Water Reactors," U. S. Nuclear Regulatory

8. W. E. Nyer, G. 0. Bright, R. J. McWhorter, Commission, Washington, D. C.

"Reactor Excursion Behavior," International Conference on the Peaceful Uses of Atomic Energy, paper 283, Geneva (1966). 21. J. M. Selby, et al., "Considerations in the Assessment of the Consequences of Efflu-

9. E. D. Clayton, "Anomalies of Criticality," ents from Mixed Oxide Fuel Fabrication Nucl. Technoj., Vol. 23, No. 14 (1974). Plants," BNWL-1697, Rev. 1 (UC-41),

Pacific Northwest Laboratories, Richland,

10. R. 0. Gumprecht, "Mathematical Basis of Washington (June 1975)..

Computer Code RIBD," DUN-4136, Douglas United Nuclear, Inc. (June 1968).

22. "Compilations of Fission Product Yields,"

11. The Effects of Nuclear Weapons, Revised NEDO-12154-1, M. E. Meek and B. F.

Edition, Samuel Glasstone, Editor, U.S. Rider, General Electric Vallecitos Nuclear Dept. of Defense (Feb. 1964). Center, TIC, P.O. Box 62, Oak Ridge, Tennessee 37830 (January 1974).

12. "Permissible Dose for Internal Radiation,"

Publication 2, Report of Committee II,

International Committee on Radiological 23. "Nuclear Decay Data for Radionuclides Protection (ICRP), Pergamon Press (1959). Occurring in Routine Releases from Nuclear Fuel Cycle Facilities," ORNL/ NUREG/TM-

13. Meteorology and Atomic Energy-- 1968, 102, D.C. Kocher, Oak Ridge National D. H. Slade, Editor, U.S. Atomic Energy Laboratory, Oak Ridge, Tennessee 37380

Commission (July 1968). (August 1977).

3.35-9

TABLE 1 RADIOACTIVITY (Ci) AND AVERAGE BETA AND GAMMA ENERGIES (MeV/dis)

OF IMPORTANT NUCLIDES RELEASED FROM CRITICALITY ACCIDENT IN THIS GUIDE

C C

Nuclide Half-life b a 0-0.5 Hr. 0.5-8 Hr. Total Y

Kr-83m 1.8 h 1. 5E+1 9.5E+1 1. 1E+2 2.6E-3 0

Kr-85m 4.5 h 9.9E0 6. 1E+I 7. IE+1 1.6E-1 2.5E-1 Kr-85 10.7 y 1. 2E-4 7.2E-4 8. 1E-4 2.2E-3 2.5E-1 Kr-87 76.3 m 6. OE+1 3.7E+2 4.3E+2 7.8E-1 1. 3E0

Kr-88 2.8 h 3'. 2E+1 2.0E+2 2.3E+2 2. OEO 3.5E-1 Kr-89 3.2 m 1.8E+3 1. 1E+4 1. 3E+4 1.6E0 1. 3E0

Xe-131m 11.9 d 1.4E-2 8.6E-2 1.OE-1 2.OE-2 1.4E-1 Xe-133m 2.0 d 3.1E-1 1. 9E0 2.2E0 4. 1E-2 1.9E-1 Xe-133 5.2 d 3.8E0 2.3E+1 2. 7E+I 4.6E-2 1.1E-1 Xe- 135m 15.6 m 4.6E+2 2.8E+3 3.3E+3 4.3E-1 9.OE-2 Xe- 135 9.1 h 5. 7E+ 1 3.5E+2 4. IE+2 2.5E-1 3.7E-1 Xe- 137 3.8 m 6.9E+3 4.2E+4 4.9E+4 1.6E-1 1. 8EO

Xe- 138 14.2 m 1.5E+3 9.5E+3 1. IE+4 1. lEO 6.2E-1

1-131 8.0 d 1.5E0 9.5E0 1. 1E+I 3.8E-1 1.9E-1

1-132 2.3 h 1.7E+2 1.OE+3 1.2E+3 2.2E0 5.OE-1

1-133 20.8 h 2.2E+l 1.4E+2 1.6E+2 6. IE-1 4. 1E-1

1-134 52.6 m 6.OE+2 3.7E+3 4.3E+3 2.6E0 6.1E-1

,1-135 6.6 h 6.3E+l 3.9E+2 4.5E+2 1.5E0 3.7E-1 Pu- 2 38 d 5.9E-4 Pu-239 2.7E-5 Pu-240 5.8E-5 Pu-241 1.8E-2 Pu-242 4.3E-7 Am-241 2.41E-5 aTotal curies, except for Pu and Am, are based on cumulative yield for fission energy spectrum using data in Ref. 22. The assumption of cumulative yield is very conservative, e.g., it does not consider appropriate decay schemes. Calculations regarding individual nuclide yields and decay schemes may be considered on an individual case basis. Data in this table does not include the iodine reduction factor allowed in Section C.2.a of this guide.

b y = year h = hour d = day m = minutes cHalf-lives and average energies derived from data in Ref. 23.

dTotal radioactivity assumes the isotopic mix to be the equilibrium mix for recycled plutonium and 1 mg of Pu 02 released (Ref. 21).

3.35-10

I. TABLE 2 VALUES OF THE CLEARANCE PARAMETERS FOR THE TASK GROUP LUNG MODELa COMPARTMENT CLASS Dbc CLASS Wc CLASS yC

d fd fd d NP Td k k k a 0.01 0.5 0.01 0.1 0.01 0.01 b 0.01 0.5 0.4 0.9 0.4 0.99 TB c 0.01 0.95 0.01 0.5 0.01 0.01 d 0.2 0.05 0.2 0.5 0.2 0.99 P e 0.5 0.8 50 0.15 500 0.05 f n.a.e n.a. 1.0 0.4 1.0 0.4 g n.a. n.a. 50 0.4 500 0.4 h 0.5 0.2 50 0.05 500 0.15 L i 0.5 1.0 50 1.0 1000 0.9

0 aSee Figure 2 for the task group lung model (TGLM) schematic diagram.

bData for soluble plutonium is included. To maintain dose conversion conservatism, this class should only be con- sidered if justified on an individual case basis.

Cclass D = readily soluble compounds where removal time is measured in days.

Class W = compounds with limited solubility where removal time is measured In weeks.

Class Y = insoluble compounds where removal time is measured in years.

dTk is the biological removal half time in days; fk is the fraction of original deposit leaving the organ via pathway indicated on the schematic model shown in Figure 2. Data are based on a mass median aerodynamic diameter of

1 micron and were developed by Battelle Memorial Institute, Pacific Northwest Laboratories, and presented in an interim report by E. C. Watson, J. R. Houston, and D. L. Strenge, April 1974.

en.a. means not applicable.

3.35-11

WI

1.0

0.0 g/cm 00

-4-4

. . . .. - /

I._ A

10 A f L0.05

10-2 S I 0.0. 012

10-3L

S0.FIGUR I

.a. . n.u. Bea E ery e RAIOO DOS TO SUFC

IET DOEAI!UCTO EAEERYSETA

fo niPln it oreo I InIntThcesadfoAlwdSptr

. .. 1.

De0e2oped frmCosdrain Prsne inRfrnc 5 hatr1

: FIGURE 1

0.1 1 3. 10

:::  : 3:.3 5-12; #

LM, CLMF L

SCHEMATIC DIAGRAM DEVELOPED FROM ICRP TASK GROUP LUNG MODEL (Refil17 FIGURE 2 I.

3.35-13

i .....- .-

. :-.' .

• -.....-; ........ .

ELEVATED RELEASE -*

• ATMOSPHERIC DISPERSION FACTORS i

. FOR FUMIGATION CONDITIONS T--

-ATMOSPHERIC CONDITIONS-

PASQUILL TYPE F

WINDSPEED 1 METERISEC

10-2

-7:i

44i

. . t . . -

ra

10 -6 Distance from Release Point (meters)

FIGURE 3 (Ref. 20)

3.35-14

10-3 I I I I I I I I II II

ELEVATED

ATMOSPHERIC DIFI

0-8 HOUR RE

U.

0r- E0

a* h =125 meters*

h = 150 meters

10-6

10-7 1 1 1 1 idl I I L -I Isall

102 103 104 Distance from Release Point (meters)

FIGURE 4 (Ref. 20)

3.35-15

ELEVATED RELEASE

- .. . ATMOSPHERIC DIFFUSION FACTORS . - I

... . 8--24 HOUR RELEASE TIME-,

L.- --

- *--

- -f--,*

-V -* - 7 -- .. ., , ,+*. .-.-

r".-,

... . . - - - - ,- - --

I I i I I

Ný

Distance from Release Point (meters)

FIGURE 5 (Ref. 20)

3.35-16

-w

3

2.5

2

0.5A = 500 meters

0.5

2

2 0.5A = 1000 meters

2

0.5A = 1500 meters

2

, 0.5A = 2000 meters

. 1.5 CoC

0.50-

102 103

104 Distance from Structure (meters)

FIGURE 6 (Ref. 20)

I

i i i i.i Ii i - i i

: *-.GROUND LEVEL RELEASE

-ATMOSPHERIC DIFFUSION FACTORS FOR

VARIOUS TIMES FOLLOWING ACCIDENT

w+-

• 0"-8 howun..-- F*--*

• I . . .. . .. . . . . . _ I.

10-

" 4--- -L - ... . . .... --- - z,- l + +] . f.

• -24 houri, i.

"-- I

4-

10-

5 . --

102 103 105 Disunce from Structure (mutre)

K

FIGURE 7A (Ref. 20)

3.35-18

Distman fo'

from 8tructure (metesr)

FIGURE 7B (Oef.20)

3.35-19

UNITED .STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 20555

0

POSTAGE AND FEESPAID

WEIED STATESNUCLEAR

OFFICIAL BUSINESS EGIRATORYCOMMISSION

PENALTY FOR PRIVATE USE,$300