Regulatory Guide 3.34

Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Uranium Fuel Fabrication Plant.

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Revision 1 U.S. NUCLEAR REGULATORY COMMISSION July 1979 REGULATORY GUIDE

OFFICE OF STANDARDS DEVELOPMENT

REGULATORY GUIDE 3.34 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL

CONSEQUENCES OF ACCIDENTAL NUCLEAR CRITICALITY IN

A URANIUM FUEL FABRICATION PLANT

B. DISCUSSION

Section 70.23, "Requirements for the ap- In the process of reviewing applications for proval of applications," of 10 CFR Part 70, licenses to operate uranium fuel fabrication

"Domestic Licensing of Special Nuclear Mate- plants, the NRC staff has developed appro- rials," requires, among other things, that the priately conservative assumptions that are used applicant's proposed equipment and facilities be by the staff to evaluate an estimate of the adequate to protect health and minimize danger radiological consequences of various postulated to life or property. In order to demonstrate the accidents. These assumptions are based on adequacy of the facility, the applicant must previous accident experience, engineering provide an analysis and evaluation of the judgment, and the analysis of applicable design and performance of structures, sys- experimental results from safety research tems, and components of the facility. The programs. This guide lists assumptions used to objective of this analysis and evaluation is to evaluate the magnitude and radiological con- assess the risk to public health and safety sequences of a criticality accident in a uranium resulting from operation of the facility, includ- fuel fabrication plant.

ing determination of the adequacy of struc- tures, systems, and components provided for A criticality accident is an accident resulting the prevention of accidents and the mitigation in the uncontrolled release of energy from an of the consequences of accidents. assemblage of fissile material. The circum- stances of a criticality accident are difficult to In a uranium fuel fabrication plant, a criti- predict. However, the most serious criticality cality accident is one of the postulated acci- accident would be expected to occur when the dents used to evaluate the adequacy of an reactivity (the extent of the deviation from applicant's proposed activities with respect to criticality of a nuclear chain reacting medium)

• public heAlth and safety. This guide describes could increase most rapidly and without control methods used by the NRC staff in the analysis in the accumulation of the largest credible of such accidentt. These methods result from mass. In a uranium fuel fabrication plant where review and action on a number of specific cases conditions that might lead to criticality are and, as such, reflect the latest general NRC- carefully avoided because of the potential for approved approaches to the problem. If an adverse physical and radiological effects, such applicant desires to employ new information an accident is extremely uncommon. However, that may be developed in the future or to use experience with these and related facilities has an alternative method, NRC will review the demonstrated that criticality accidents could proposal and approve its use, if found occur.

acceptable.

In a uranium fuel fabrication plant, such an accident might be initiated by (1) inadvertent Lines indicate substantive changes from previous issue. transfer or leakage of a solution of fissile USNRC REGULATORY GUIDES Comments should be sent to the Secretary of the Commission, U.S. Nuclear RegulatOry Commission, Washington, D.C. 20555, 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:

ating 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 pliance with them is not required. Methods and solutions different from those 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 these 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 divisions 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.

material from a geometrically safe containing mechanism may be counteracted, the initial vessel into an area or vessel not so designed, burst was frequently succeeded by a plateau

(2) introduction of excess fissile material solu- period of varying length. This plateau was tion to a vessel, (3) introduction of excess characterized by a lesser and declining fission fissile material to a solution, (4) overconcen- rate and finally by a further dropoff as shut- tration of a solution, (5) failure to maintain down was completed. The magnitude of the sufficient neutron absorbing materials in a initial burst was directly related to the rate of vessel, (6) precipitation of fissile solids from a increase of reactivity and its magnitude above solution and their retention in a vessel, the just-critical value but was inversely related

(7) introduction of neutron moderators or to the background neutron flux.

reflectors (e.g., by addition of water to a highly undermoderated system), (8) deforma- Those systems consisting only of solid fis- tion of or failure to maintain safe storage sile, reflector, or moderator materials exhibited arrays, or (9) similar actions that can lead to little or no plateau period, whereas solution increases in the reactivity of fissile systems. systems had well developed plateaus. For solu- Some acceptable means for minimizing the likeli- tion systems, the energy release during the hood of such accidents are described in Regu- plateau period, because of its duration, pro- latory Guides 3.4, "Nuclear Criticality Safety vided the major portion of the total energy in Operations with Fissionable Material Outside released. For purposes of the planning neces- Reactors,"' and 3.1, "Use of Borosilicate Glass sary to deal adequately with criticality Raschig Rings as a Neutron Absorber in incidents in experimental and production-type Solutions of Fissile Material."' nuclear facilities, Woodcock (Ref. 2) made use of these data to estimate possible fission yields

1. CRITICALITY ACCIDENT EXPERIENCE IN RELATION TO from excursions in various types of systems.

THE ESTIMATION OF THE MOST SEVERE ACCIDENT For example, spike yields of lE+17 and 1E+18 and total yields of 3E+18 and 3E+19 fissions Stratton (Ref. 1) has reviewed in detail were suggested for criticality accidents

34 occasions prior to 1966 when the power level occurring in solution systems of 100 gallons or of a fissile system increased without control as less and more than 100 gallons, respectively.

a result of unplanned or unexpected changes in Little or no mechanical damage was predicted at its reactivity. Although only six of these these levels.

occurred in processing operations, and the remainder occurred mostly in facilities for

2. METHODS DEVELOPED FOR PREDICTING THE MAGNITUDE

obtaining criticality data or in experimental OF CRITICALITY ACCIDENTS

reactors, the information obtained and its correlation with the characteristics of each The nuclear excursion behavior of solu- system have been of considerable value for use tions of enriched uranium has been studied in estimating the consequences of accidental extensively both theoretically and experi- criticality in process systems. The incidents mentally. Dunenfeld and Stitt (Ref. 3)

occurred in aqueous solutions of uranium or summarize the kinetic experiments on water plutonium (10), in metallic uranium or boilers using uranyl sulfate solutions and plutonium in air (9), in inhomogeneous water- describe the development of a kinetic model moderated systems (9), and in miscellaneous that was confirmed by experiment. This model solid uranium systems (6). The estimated total defines the effects of thermal expansion and number of fissions per incident ranged from radiolytic gas formation as power-limiting and lE+15 2 to 1E+20 with a median of about 2E+17.

shutdown mechanisms.

In ten cases, the supercriticality was halted by an automatic control device. In the remainder, The results of a series of criticality excur- the shutdown was effected as a consequence of sion experiments resulting from the introduc- the fission energy release that resulted in tion of uranyl nitrate solutions to vertical thermal expansion, density reduction from the cylindrical tanks at varying rates are formation of very small bubbles, mixing of light summarized by L4corch6 and Seale (Ref. 4).

and dense layers, loss of water moderator by This report confirms the applicability of the boiling, or expulsion of part of the mass. kinetics model for solutions, provides correla- tions of peak power with reactivity addition Generally, the criticality incidents were rate, notes the importance of a strong neutron characterized by an initial burst or spike in a curve of fission rate versus time followed by a source in limiting peak power, and indicates the nature of the plateau following the peak.

rapid but incomplete decay as the shutoff mechanism was initiated. As more than one Many operations with fissile materials in a shutdown mechanism may affect the reactivity uranium fuel fabrication plant are conducted of the system and the effect of a particular with aqueous (or organic solvent) solutions of fissile materials. Consequently, well-founded

'Copies may be obtained from the U.S. Nuclear Regulatory methods for the prediction of total fissions and Commission, Washington, D.C. 20555, Attention: Director, maximum fission rate for accidents that might Division of Document control. occur in solutions (in process or other vessels)

2

1E+15 = I x 10's. This notational form will be used through- by the addition of fissile materials should be of out this guide. considerable value in evaluating the effects of

3.34-2

possible fabrication plant criticality accidents. of containers is maintained. Normally only a From the results of excursion studies and from limited number of containers may be in motion accident data, Tuck (Ref. 5) has developed in the vicinity of the array. Consequently, the methods for estimating (1) the maximum num- rate of reactivity addition in such a system ber of fissions in a 5-second interval (the first would be lower, and the predicted magnitude of spike), (2) the total number of fissions, and criticality incident would be correspondingly

(3) the maximum specific fission rate in lower.

vertical cylindrical vessels, 28 to 152 cm in diameter and separated by >30 cm from a For systems other than solutions systems, bottom reflecting surface, resulting from the the estimation of the peak fission rate and the addition of up to 500 g/1l solutions of Pu-239 or total number of fissions accompanying an U-235 to the vessel at rates of 0.7 to accidental nuclear criticality may also be esti-

7.5 gal/min. Tuck also gives a method for mated with the aid of information derived from estimating the power level from which the accident experience, from experiments on steam-generated pressure may be calculated reactors utilizing bare uranium metal (Ref. 7),

and indicates that use of the formulas for tanks and from the SPERT-.1 reactor transient tests

>152 cm in diameter is possible with a loss in with light-water and heavy-water moderated accuracy. uranium-aluminum and U0 2 -stainless steel fuels (Ref. 8). Oxide core tests in the latter group Methods for estimating the number of fis- provide some information on energy release sions in the initial burst and the total number mechanisms that may be effective, for example, of fissions, derived from the work reported by in fuel storage areas. Review of unusual pro- L6corch6 and Seale (Ref. 4), have also been cessing structures, systems, and components developed by Olsen and others (Ref. 6). These for the possibility of accidental criticality were evaluated by application to ten actual should also consider recognized anomalous accidents that have occurred in solutions and situations in which the possibility of accidental were shown to give conservative estimates in nuclear criticality may be conceived (Ref. 9).

all cases except one.

The application of the double-contingency Fission yields for criticality accidents principle3 to fissile material processing opera- occurring in solutions and some heterogeneous tions has been successful in reducing the systems, e.g., aqueous/fixed geometry, can be probability of accidental criticality to a low estimated with reasonable accuracy using value. As a consequence, the scenarios re- existing methods. However, methods for quired to arrive at accidental criticality involve estimating possible fission yield from other the assumption of multiple breakdowns in the types of heterogeneous systems, e.g., aque- nuclear criticality safety controls. It has ous/powder, are less reliable because of the therefore been a practice to simply and con- uncertainties involved in predicting the reac- servatively assume an accidental criticality of a tivity rate. The uncertainty of geometry and magnitude equal to, or some multiple of, the moderation results in a broad range of possible historical maximum for all criticality accidents yields. outside reactors without using any scenario clearly defined by the specific operations being Woodcock (Ref. 2) estimated that in solid evaluated. In the absence of sufficient plutonium systems, solid uranium systems, and guidance, there has been wide variation in the heterogeneous liquid/powder systems (fissile credibility of the postulated magnitude of the material not specified) total fission yields occurrence (particularly the size of the initial (substantially occurring with the spike) of burst), the amount of energy and radioactivity

1E+18, 3E+19, and 3E+20, respectively, could assumed to be released, and the magnitude of be predicted. Mechanical damage varied from the calculated consequences.

slight to extensive. Heterogeneous systems consisting of metals or solids in water were It is the staff's judgment that the evalua- estimated to achieve a possible magnitude of tion of the criticality accident should assume

1E+19 following an initial burst of 3E+18 the simultaneous breakdown of at least two fissions. The possibility of a larger fission independent controls throughout all elements of burst (possibly as high as 3E+22) resulting in the operation. Each control should be such that a serious explosion could be conceived for its circumvention is of very low probability.

large storage arrays where prompt criticality Experience has shown that the simultaneous was exceeded, e.g., by collapse of shelving. It failure of two independent controls is very is recognized that in such arrays, where unlikely if the controls are derived, applied, reactivity is more likely to be increased by the and maintained with a high level of quality successive additions of small increments of assurance. However, if controls highly material, only a delayed critical condition with dependent on human actions are involved, this maximum yields of 1E+19 fissions is likely. approach will call for some variation in the These estimates could aid in the analysis of assumed number of control failures. The situations in plant systems. However, they 3 should not be taken as absolute values for The double-contingency principle is defined in ANSI N16.1-

1975, "Nuclear Criticality Safety in Operations with Fissionable criticality assumptions for the purpose of this Materials Outside Reactor," which is endorsed by Regulatory guide. In a product storage area, a rigid array Guide 3.4.

3.34-3

criticality accidents so conceived should then introduced. To determine the circumstances of be analyzed to determine the most severe the criticality accidents, controls judged within the framework of assumed control equivalent to at least two highly reliable, failures, using realistic values of such independent criticality controls should be variables as the fissile inventory, vessel sizes, assumed to be circumvented. The magnitude of and pump transfer rates. the possible accidents should then be assessed, on an individual case basis, to estimate the

3. RADIOLOGICAL CONSEQUENCES OF ACCIDENTAL CRITI- extent and nature of possible effects and to CALITY provide source terms for dose calculations. The most severe accident should then be selected Past practice has been to evaluate the ra- for the assessment of the adequacy of the diological consequences to individuals of postu- plant.

lated accidental criticality in uranium fuel fabrication plants in terms of a fraction of the Calculation of the radioactivity of sig- guideline values in 10 CFR Part 100, "Reactor nificant fission products produced in the Site Criteria." excursion may be accomplished using the com- puter code RIBD (Ref. 10). An equivalent cal- The consequences of a criticality accident culation may be substituted, if justified on an may be limited by containment, shielding, iso- individual case basis.

lation distance, or evacuation of adjacent occupied areas subsequent to detection of the b. If the results of the preceding evalua- accident. If the impact of a criticality accident tion indicate that no possible criticality is to be limited through evacuation of adjacent accident exceeds in severity the criticality occupied areas, there should be prior, formal accident postulated in this section, then the arrangements with individual occupants and conditions of the following example may be local authorities sufficient to ensure that such assumed for the purpose of assessing the movements can be effected in the time allowed. adequacy of the facility. A less conservative set of conditions may be used if they are shown The equations provided for estimating to be applicable by the specific analyses con- doses from prompt gamma and neutron radiation ducted in accordance with paragraph C.1.a were developed using experimental and above.

historical data. The report, "Prompt Neutron and Gamma Doses from an Accidental Critical- An excursion is assumed to occur in a ity," explains this development.* These equa- vented vessel of unfavorable geometry con- tions cannot be expected to be as accurate as taining a solution of 400 g/1 of uranium detailed calculations based on actual accident enriched in U-235. The excursion produces an conditions. Comparisons with published in- initial burst of lE+18 fissions in 0.5 second formation indicate they may not be conservative followed successively at 10-minute intervals by for smaller accidents (e.g., 1-2E+17 fissions). 47 bursts of 1.9-E+17 fissions for a total of However, for accidents that are likely to be 1E+19 fissions in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The excursion is assumed for safety assessment purposes, they assumed to be terminated by evaporation of appear to be sufficiently conservative. These 100 liters of the solutions.

equations are included in the guide to provide a simplified method for estimating prompt gamma and neutron doses from a potential 2. ASSUMPTIONS RELATED TO THE RELEASE 4 OF RADIO-

criticality accident. ACTIVE MATERIAL ARE AS FOLLOWS:

C. REGULATORY POSITION

a. It should be assumed that all of the noble gas fission products and 25% of the iodine

1. FOLLOWING ARE THE PLANT ASSESSMENT AND ASSUMP- radionuclides resulting from the excursion are TIONS RELATED TO ENERGY RELEASE FROM A CRITI- released directly to a ventilated room atmos- CALITY ACCIDENT AND THE MINIMUM CRITICALITY phere. It should also be assumed that an ACCIDENT TO BE CONSIDERED: aerosol, which is generated from the evapora- tion of solution during the excursion, is a. When defining the characteristics of an released directly to the room atmosphere. The assumed criticality accident in order to assess aerosol should be assumed to comprise 0.05% of the adequacy of structures, systems, and com- the salt content of the solution that is ponents provided for the .prevention or evaporated. The room volume and air ventila- mitigation of the consequences of accidents, tion rate and retention time should be con- the applicant should evaluate credible critical- sidered on an individual case basis.

ity accidents in all those elements of the plant provided for the storage, handling, or pro- b. The effects of radiological decay during cessing of fissile materials or into which fissile transit within the plant should be evaluated on materials in significant amounts could be an individual case basis.

A copy of Charles A. Willis' report, "Prompt Neutron and 4 Gamma Doses from an Accidental Criticality," is available for Certain 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 posi- NW., Washington, D.C. tion indicated in Regulatory Guide 1.3 (Ref. 18).

3.34-4

c. A reduction in the amount of radioactive For concrete, the dose should be material available for release to the plant reduced by a factor of 2.3 for the first 8 environment through filtration systems in the inches, 4.6 for the first foot, and a factor of

1plant exhaust system(s) may be taken into 20 for each additional foot.

account, but the amount of reduction in the concentration of radioactive materials should be b. No correction should be made for deple- evaluated on an individual case base. tion of radioactive iodine from the effluent plume due to deposition on the ground or for d. Table 1 lists the radioactivity of sig- the radiological decay of iodine in transit.

nificant radionuclides released, but it does not include the iodine depletion allowance. 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 rate of a person offsite should be assumed to

3. ACCEPTABLE ASSUMPTIONS FOR DOSE AND DOSE CON- be 3.47E-4m3 /sec. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following VERSION ARE AS FOLLOWS: the accident, the breathing rate should be as- sumed to be 1.75E-4m3 /sec. These values were a. The applicant should show that the con- developed from the average daily breathing sequences of the prompt gamma and neutron rate (2E+7 cm 3 /day) assumed in the report of dose are sufficiently mitigated to allow ICRP Committee 11-1959 (Ref. 12).

occupancy of areas necessary to maintain the plant in a safe condition following the accident. d. External whole body doses should be The applicant should estimate the prompt calculated using "infinite cloud" assumptions, gamma and neutron dose that could be received i.e., the dimensions of the cloud are assumed at the closest site boundary and nearest to be large compared to the distance that the residence. The following semi-empirical equa- gamma rays and beta particles travel. "Such a tions may be used for these calculations. cloud would be considered an infinite cloud for Because detailed evaluations will be dependent a receptor at the center because any additional on the site and plant design, different methods [gamma and] beta emitting material beyond the may be substituted on an individual case basis. cloud dimensions would not alter the flux of Potential dose attenuation due to shielding and [gamma rays and] beta particles to the dose exposures should be evaluated on an indi- receptor." [See Meteorology and Atomic vidual case basis. Energy--1968 (Ref. 13), Section 7.4.1.1; edi-

(1) Prompt 5 Gamma Dose torial additions made so that gamma and beta emitting material could be considered.] Under these conditions the rate of energy absorption D = 2.1E-20 N d-2e-3.4d Y per unit volume is equal to the rate of energy where released per unit volume. For an infinite uniform cloud containing X curies of beta radio- D Y = gamma dose (rem) activity per cubic meter, the beta dose rate in air at the cloud center is N = number of fissions D- - 0.475E Px d = distance from source (kin).

Data presented in The Effects of Nuclear The surface body dose rate from beta emitters in the infinite cloud can be approximated as being Weapons (Ref. 11, p. 384) may be used to one-half this amount (i.e., pDIo = 0.23Epx).

develop dose reduction factors. For concrete, For gamma emitting material, the dose rate in the dose should be reduced by a factor of 2.5 for the first 8 inches, a factor of 5.0 for the air at the cloud center is first foot, and a factor of 5.5 for each addi- tional foot. D- = 0.507Eyx

(2) Prompt Neutron Dose From a semi-infinite cloud, the gamma dose rate in air is Dn = 7E-20N d-2e-s.2d where D-s = 0.25E X

Dn = neutron dose (rem) Y Y

N = number of fissions where d = distance from source (kin).

D- = beta dose rate from an infinite

5Most of the gamma radiation is emitted in the actual fission cloud (rad/sec)

process. Some gamma radiation is produced in various secondary nuclear processes, including decay of fission Des = gamma dose rate from an infinite products. For the purposes of this guide, "prompt" gamma Y cloud (rad/sec)

doses should be 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 E = average beta energy per disintegration given includes these considerations. (MeV/dis)

3.34-5

E = average gamma energy per disintegration 4. ACCEPTABLE ASSUMPTIONS FOR ATMOSPHERIC DIFFU-

(MeV/dis) SION ARE AS FOLLOWS:

X = concentration of beta or gamma emitting a. If the uranium fuel fabrication plant

3 isotope in the cloud (Ci/m ). gaseous effluents are exhausted through a stack, the assumptions presented in Regulatory e. The following specific assumptions are Guide 3.35, "Assumptions Used for Evaluating acceptable with respect to the radioactive cloud the Potential Radiological Consequences of dose calculations: Accidental Nuclear Criticality in a Plutonium Processing and Fuel Fabrication Plant,"

(1) The dose at any distance from the Regulatory Positions C.4.a and C.4.b should plant should be calculated based on the maxi- be used to calculate the atmospheric diffusion mum concentration time integral (in the course factors.

• of the accident) in the plume at that distance, taking into account specific meteorological, b. If no onsite meteorological data are topographical, and other characteristics that available for facilities exhausted without may affect the maximum plume concentration. stacks, the atmospheric diffusion model should These site-related characteristics should be be as follows:

evaluated on an individual case basis. In the case of beta radiation, the receptor is assumed (1) The 0-to-8 hour ground level release to be exposed to an infinite cloud at the concentrations may be reduced by a factor maximum ground level concentration at that ranging from one to a maximum of three (see distance from the plant. In the case of gamma Fig. 2) for additional dispersion produced by radiation, the receptor is assumed to be the turbulent wake of a major building in exposed to only one-half the cloud owing to the calculating nearby potential exposures. The presence of the ground. The maximum cloud volumetric building wake correction factor, as concentration should always be assumed to be defined in Section 3.3.5.2 of Meteorology and at ground level. Atomic Energy--1968 (Ref. 13), should be used in the 0-to-8 hour period only; it is used with

(2) The appropriate average beta and a shape factor of one-half and the minimum gamma energies emitted per disintegration may cross-sectional area of a major building only.

be derived from the Table of Isotopes (Ref. 14) or other appropriate sources, e.g., (2) The basic equation for atmospheric Ref. 21. diffusion from a ground level point source is

(3) The whole body dose should be con- 1 x/Q- - uoa sidered as the dose from gamma radiation at a yz depth of 5 cm and the genetic dose at a depth of 1 cm. The skin dose should be the sum of where the surface gamma dose and the beta dose at a depth of 7 mg/cm 2 . The beta skin dose may be x = the short-term average centerline value estimated by applying an energy-dependent of the ground level concentration attenuation factor (Dd/DB) to the surface dose (Ci/m 3 )

according to a method developed by Loevinger, Japha, and Brownell (Ref. 15). See Figure 1.

Q = amount of material release (Ci/sec)

f. The "critical organ" dose from the inhaled u = windspeed (m/sec)

radioactive materials should be estimated. The

"critical organ" is that organ that receives the a = the horizontal standard deviation of the y plume (m). [See Ref. 17, Figure V-1, highest radiation dose after the isotope is absorbed into the body. For the purpose of p. 48.]

this guide, the following assumptions should be made: o = the vertical standard deviation of the z plume (m). [See Ref. 17, Figure V-2,

(1) The radionuclide dose conversion p. 48.]

factors are as recommended by the report of Committee II, ICRP (1959) (Ref. 12) or other (3) For time periods of greater than appropriate source. 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 meander and spread uniformly over a 22.50

6

(2) The effective half-life for the nuclide sector. The resultant equation is is as recommended in ICRP Publication 6 (Ref. 2.032

16) or other appropriate source. x/Q -ux g. The potential dose for all significant

"The 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 nuclides should be estimated for the population meteorological data are available to justify a wind direction distribution on a site-related basis. change. This should be considered on an individual case basis.

3.34-6

where (5) Figures 3A and 3B give the ground level release atmospheric diffusion factors x = distance from point of release to the re- based on the parameters given in b(4).

ceptor; other variables are given in b(2).

(4) The atmospheric diffusion model 7 for

D. IMPLEMENTATION

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

Accident Atmospheric Conditions Except in those cases in which the applicant Pasquill Type F; windspeed 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 /> 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.

7

1n some cases site-dependent parameters such as meteo- If an applicant wishes to use this regulatory rology, topography, and local geography may dictate the use of guide in developing submittals for applications a more restrictive model to ensure a conservative estimate of docketed on or before December 1, 1977, the potential offsite exposures. In such cases, appropriate site- related meteorology should be developed on an individual case pertinent portions of the application will be basis. evaluated on the basis of this guide.

3.34-7

REFERENCES

1. W. R. Stratton, "Review of Criticality 12. "Permissible Dose for Internal Radiation,"

Incidents," LA-3611, Los Alamos Scientifc Publication 2, Report of Committee II,

Laboratory (Jan. 1967). International Commission on Radiological Protection, Pergamon Press (1959).

2. E. R. Woodcock, "Potential Magnitude of Criticality Accidents," AHSB(RP)R-14, 13. Meteorology and Atomic Energy--1968, United Kingdom Atomic Energy Authority. D. H. Slade, Editor, U.S. Atomic Energy Commission (July 1968).

3. M. S. Dunenfeld, R. K. Stitt, "Summary Review of the Kinetics Experiments on 14. C. M. Lederer, J. M. Hollander, I. Perl- Water Boilers," NAA-SR-7087, Atomic man, Table of Isotopes, 6th Ed., Lawrence International (Feb. 1973). Radiation Laboratory, Univ. of California, Berkeley, CA (1967).

4. P. L~corch6, R. L. Seale, "A Review of the Experiments Performed to Determine 15. Radiation Dosimetry, G. J. Hine and G. L.

the Radiological Consequences of a Criti- Brownell, Editors, Academic Press, New cality Accident," Y-CDC-12, Union Carbide York (1956)

Corp. (Nov. 1973).

16. Recommendations of ICRP, Publication 6,

5. G. Tuck, "Simplified Methods of Estimating Pergamon Press (1962).

the Results of Accidental Solution Excursions," Nucl. Technol., Vol. 23, 17. F. A. Gifford, Jr., "Use of Routine Meteor- p. 177 (1974). ological Observations for Estimating Atmos- pheric Dispersion," Nuclear Safety, Vol. 2,

6. A. R. Olsen, R. L. Hooper, V. 0. Uotinen, No. 4, p. 48 (June 1961).

C. L. Brown, "Empirical Model to Estimate Energy Release from Accidental Criticality," 18. Regulatory Guide 1.3, "Assumptions Used ANS Trans., Vol. 19, p. 189-91 (1974). for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident

7. T. F. Wimmette et al., "Godiva 2--An for Boiling Water Reactors," U.S. Nuclear Unmoderated Pulse Irradiation Reactor," Regulatory Commission, Washington, D.C.

Nucl. Sci. Eng., Vol. 8, p. 691 (1960). (June 1974).

8. W. E. Nyer, G. 0. Bright, R. J. 19. "Radiological Health Handbook, " U.S.

McWhorter, "Reactor Excursion Behavior," Department of Health, Education and Wel- International Conference on the Peaceful fare (January 1970).

Uses of Atomic Energy, paper 283, Geneva

(1966). 20. "Compilation of Fission Product Yields,"

NEDO-12154-1, M. E. Meek and B. F.

9. E. D. Clayton, "Anomalies of Criticality," Rider, General Electric Vallecitos Nuclear Nucl. Technol., Vol. 23, No. 14(1974). Center, TIC, P.O. Box 62, Oak Ridge, Tennessee 37830 (January 1974).

10. R. 0. Gumprecht, "Mathematical Basis of Computer Code RIBD," DUN-4136, Douglas 21. "Nuclear Decay Data for Radionuclides Oc- United Nuclear, Inc. (June 1968). curring in Routine Releases from Nuclear Fuel Cycle Facilities," ORNL/NUREG/TM-

11. The Effects of Nuclear Weapons, Revised 102, D. C. Kocher, Oak Ridge National Edition, S. Glasstone, Editor, U.S. Dept. Laboratory, Oak Ridge, Tennessee 37830

of Defense (1964). (August 1977).

3.34-8

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

OF IMPORTANT NUCLIDES RELEASED FROM CRITICALITY ACCIDENT IN THIS GUIDE

Radioactivitya Nuclide Half-lifeb, c 0-0.5 Hr. 0.5-8 H

r. Total b

Kr-83m 1.8 h 2.2E+1 1.4E+2 1.6E+2 2.6E-3 0

Kr-85m 4.5 hr 2. IE+I 1.3E+2 1.5E+2 1.6E-1 2.5E-I

Kr-85 10.7 y 2.2E-4 1.4E-3 1.6E-3 2.2E-3 2.5E-I

Kr-87 76.3 m 1.4E+2 8.5E+2 9.9E+2 7.8E-I 1.3E0

Kr-88 2.8 h 9. IE+I 5.6E+2 6.5E+2 2.0E0 3.5E-i Kr-89 3.2 m 5.9E+3 3.6E+4 4.2E+4 1.6E0 i. 3E0

Xe-131m 11.9 d 1.1E-2 7. 0E-2 8.2E-2 2. 0E-2 1.4E-1 Xe- 133m 2.0 d 2.5E-1 1.6E0 1.8E0 .4.1E-2 1.9E-2 Xe-133 5.2 d 3.8E0 2.3E+1 2.7E+1 4.6E-2 1. IE-1 Xe- 135m 15.6 m 3. IE+2 1.9E+3 2.2E+3 4.3E-2 9.0E-2 Xe-135 9.1 h 5.0E+i 3. IE+2 3.6E+2 2.5E-I 3.7E-1 Xe-137 3.8 m 6.9E+3 4. 2E+4 4.9E+4 1.6E-1 1.8E0

Xe-138 14.2 1.8E+3 1. 1E+4 1.3E+4 1.1E0 6.2E-1

1-131 8.0 d 1.2E0 7.5EO 8.7E0 3.8E-1 1.9E-1

1-132 2.3 h 1.5E+2 9.5E+2 1.IE+3 2.2E0 5.0E-1

1-133 20.8 h 2.2E+1 1 .4E+2 1.6E+ 2 6. 1E- 1 4. iE-1

1-134 52.6 m 6.3E+2 3.9E+3 4.5E+3 2.6E0 6.1E-1

1-135 6.6 h 6.6E+1 4.OE+2 4.7E+2 1.5EO 3.7E-1 aTotal curies are based on cumulative yields for fission energy spectrum using the data in Ref. 20. The assump- tion 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.

bllalf-lives and average energies are derived using the data in Ref. 21.

cy = year d = day h = hour m = minute

3.34-9

1.0

11 if 11 L;-4-

! I I 1 114--1 1 1 1 "1-]  ;"1 1 'IT

III I,- /11 VI

0.007 g/cm 1 -

/if

01.02

10-1 AII

~~0.2.

10--2

- - - -- -

is

ý

0.5 t

10--3

0.

1 I HI! /III 1.0

I

10

Maximum Beta Energy, MeV

1 RATIO OF DEPTH DOSE TO SURFACE DOSE AS A FUNCTION BETA ENERGY SPECTRA

for Infinite Plane Source of Infinite Thickness and for Allowed Spectra Developed from Considerations Presented in Reference 15, Chapter 16 FIGURE 1

3.34-10

3

2.5

2

0

0

0

1.5 C-)

1

0.5

0

10 2 103 104 Distance from Structure (meters)

FIGURE 2 (Ref. 18)

10-2 GROUND LEVEL RELEASE

ATMOSPHERIC DIFFUSION FACTORS

FOR VARIOUS TIMES FOLLOWING ACCIDENT

hours

10-3 E

a

0 8-24 hours

10-4 l

• J J v* I "* J I i J

10-5 1 1 1 1 1 1 a

4

102 103 10 105 Distance from Structure (meters)

FIGURE 3A (Ref. 18)

3.34-12

Distance from Structure (meters)

FIGURE 3B (Ref. 18)

3.34-13

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