Revision I U.S. NUCLEAR REGULATORY COMMISSION July 1979

'PA' S t REGULATORY GUIDE "** OFFICE OF STANDARDS DEVELOPMENT

REGULATORY GUIDE 334

ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF ACCIDENTAL NUCLEAR CRITICALITY IN A FUEL FABRICATION PLANT

A. INTRODUCTION 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 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 accidentS. 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 chances from previous issue. transfer or leakage of a solution of fissile

USNRC REGULATORY GUIDES SComrments RGULTOYiGsRguly mitcosdn eobe sent wasnton. to " Secretary D.C. of2050 toe Comrn.iokn,Attention: Docketing US. Nuclear and Regulatory Guides am issued to describe Wnd hlae avalable Wotie public Service Dranci. methods ecceptable to ta NRC staff of inplernentsng spc•pfic parts of tie Comnmision's ugulations. to delneata tecahiques used by ft staff in evai.u- The gukdes are ssued in "etfollowing t broad divisons: ~~¶fcproblerns or postulated accidents. or 00 provide guidanca 10 Regukxy Guides we not substf••es for regulation, amndCorn- 1. PWm ReactorS 6. Products pliarce with diem k not required. Methods and solutksodfferentfhose 1uorn 2. Reserch and Test Reactao 7. Transporation Wt out In eOguds vin be acptable Iftoy provide a bashsfor f 3. Fuels and Materats Faciities . Occupational Health rquisite ID die Issuance or continuance Of a permit or license by toe 4. Enwonmentst and Siting S Antitrust and Financial Review Mataerals and Plant Protection 10. General SS, Conet an~d stigestion for limprvemewit inminse Wleu weariecouraed at ptacernentIiuqua for an single an SUMMW'a copies ofdsvibA- isued guides rw fowlwhch=Ve niy copies be reproduced) of fMur Wodeor for ail trne, and guides WO be revised, =appropriate.,,soccornmodate c.vnmens in spacft, divisions -hotd be mdeIn writing to t'e US. Nucear Regulatory 4W to 1`0l new information or expiene. This guide was revised ,a result Cornissicn. Washngton, D.C. 205. Attrntion: Director, Division of at substantive conirnentUecaived from te pubic and additional stiff 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 absorbing materials in a initial burst was directly related to the rate vessel. (6) precipitation of 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, latory Guides pro 3.4, ."Nuclear Criticality Safety vided the major portion of the total in Operations energy with Fissionable Material Outside released. For purposes of the planning Reactors,"' neces and 3.1, "Use of Borosilicate Glass sary to deal adequately with Raschig criticality 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 1. CRITICALITY ACCIDENT data to estimate possible fission yields EXPERIENCE IN RELATION TO from excursions THE ESTIMATION OF THE MOST SEVERE ACCIDENT in various types of systems. For example, spike yields of IE÷i7 and 1E÷18 Stratton (Ref. 1) has reviewed in detail and total yields of 3E+18 and 3E+19 fissions 34 occasions prior, to 1966 when the power level were suggested for criticality accidents of a fissile system increased without control as occurring in solution systems of 100 gallons or a result of unplanned or unexpected changes in less and more than 100 gallons, respectively. its reactivity. Although only six of these Little or no mechanical damage was predicted at occurred in processing operations, and the these levels. remainder occurred mostly in facilities for obtaining criticality data or in experimental 2. METHODS DEVELOPED FOR PREDICTING THE MAGNITUDE reactors, the information obtained and its OF CRITICALITY ACCIDENTS correlation with the characteristics of each system have been of considerable value for use The nuclear excursion behavior of solu in estimating the consequences of accidental tions of enriched uranium has been studied criticality in process systems. The incidents extensively both theoretically and experi occurred in aqueous solutions of uranium or mentally. Dunenfeld and Stitt (Ref. 3) plutonium (10), in metallic uranium or summarize the kinetic experiments on water plutonium in air (9), in inhomogeneous water boilers using uranyl- sulfate solutions and moderated systems (9), and in miscellaneous describe the development of a kinetic model solid uranium systems (6). The estimated total that was confirmed by experiment. This model number of fissions per incident ranged defines the effects of thermal expansion and 2 from IE+15 to IE+20 with a median of about 2E÷17. radiolytic gas formation as power-limiting and In ten cases, the supercriticality was halted by shutdown mechanisms. an automatic control device. In the remainder, the shutdown was effected as a consequence of The results of a series of criticality excur the fission energy release that resulted in sion experiments resulting from the introduc thermal expansion, density reduction from the tion of uranyl nitrate solutions to vertical formation of very small bubbles, mixing of light cylindrical tanks at varying rates are and dense layers, loss of water moderator by summarized by Licorch6 and Seale (Ref. 4). boiling, or expulsion of part of the mass. This report confirms the applicability of the kinetics model for solutions, provides correla Generally, the criticality incidents were tions of peak power with reactivity addition characterized by an initial burst or spike in a rate, notes the importance of a strong neutron curve of fission rate versus time followed by a source in limiting peak power, and indicates rapid but incomplete decay as the shutoff the nature of the plateau following the peak. mechanism was initiated. As more than one shutdown mechanism may affect the reactivity Many operations with fissile materials in a of the system and the effect of a particular uranium fuel fabrication plant are conducted 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 Division of Document Control. rate for accidents that might occur in solutions (in process or other vessels) S1E'15 - 1 x 101s. This notational form will be usecrthrough by the addition of fissile materials out this guide. should be of considerable value in evaluating the effects of

3.34-2 possible fabrication plant criticality accidents. of containers is maintained. Normally only a in motion From the results of excursion studies and from limited number of containers may be in the vicinity of' the array. Consequently, the accident data, Tuck (Ref. 5) has developed rate of reactivity addition in such a system methods for estimating (1) the maximum num would be lower, and the predicted magnitude of fissions in a 5-second interval (the first ber of incident would be correspondingly (2) the total number of fissions, and criticality spike), lower. (3) the maximum specific fission rate in vertical cylindrical vessels, 28 to 152 cm in systems other than solutions systems, diameter and separated by >30 cm from a For rate and the bottom reflecting surface, resulting from the the estimation of the peak fission fissions accompanying an addition of up to 500 g/1 solutions of Pu-239 or total number of nuclear criticality may also be esti U-235 to the vessel at rates of 0.7 to accidental information derived from 7.5 gal/min. Tuck also gives, a method for mated with the aid of experience, from experiments on estimating the power level from which the accident bare uranium metal (Ref. 7), steam-generated pressure may be calculated reactors utilizing transient tests and indicates that use of the formulas for tanks and from the SPERT-l reactor >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 release Methods for estimating the number of fis provide some information on energy sions in the initial burst and the total number mechanisms that may be effective, for example, pro of fissions, derived from the work reported by in fuel storage areas. Review of unusual systems, and components LUcorchi and Seale (Ref. 4), have also been cessing structures, 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 3 Fission yields for criticality accidents principle 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 re estimated with reasonable accuracy using value. As a consequence, the scenarios involve existing methods. However, methods for quired to arrive at accidental criticality estimating possible fission yield from other the assumption of multiple breakdowns in the has types of heterogeneous systems, e.g., aque nuclear criticality safety controls. It con ous/powder, are less reliable because of the therefore been a practice to simply and 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 variation in the maximum yields of IE+19 fissions is likely. -approach will call for some assumed number of control failures: The These estimates could aid in the analysis of situations in, plant systems. However, they 3 values for The double-contingency principle is defined in ANSI N16.1 should not be taken as absolute 1975. "Nuclear criticality Safety in Operatzons 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 be analyzed the circumstances of to determine the most severe the criticality accidents, within controls judged the framework of assumed control equivalent to at least failures, using two highly reliable, realistic values of such independent criticality contiols variables as the fissile inventory, should be vessel sizes, assumed to be circumvented. The magnitude and pump transfer rates. of the possible accidents should then be assessed, on an individual 3. RADIOLOGICAL case basis, to estimate the CONSEQUENCES OF ACCIDENTAL CRITI extent and nature CALITY of possible effects and to provide source terms for dose calculations. The most severe Past practice accident should then be selected has been to evaluate the ra for diological consequences the assessment of the adequacy of the to individuals of postu plant. lated accidental criticality in uranium fuel fabrication plants in terms of a fraction of the Calculation guideline values in 10 CFR of the radioactivity of sig Part 100, "Reactor nificant fission Site Criteria." products produced in the excursion may be accomplished using the com puter code The consequences of RIBD (Ref. 10). An equivalent cal a criticality accident culation may may be limited by containment, be substituted, if justified on an shielding, iso individual case basis. lation distance, or evacuation of adjacent occupied areas subsequent to detection of the b. accident. If the impact If the results of the preceding evalua of a criticality accident tion is to be limited through indicate that no possible criticality evacuation of adjacent accident occupied areas, there exceeds in severity the criticality should be prior, formal accident arrangements with postulated in this section, then the individual occupants and conditions local authorities sufficient to ensure of the following example may be that such assumed for the movements can be effected in the time allowed. purpose of assessing the adequacy of the facility. A less conservative The set of conditions may be used if they are shown equations provided for estimating to doses from prompt gamma be applicable by the specific analyses con and neutron radiation ducted were developed using in accordance with paragraph C.1.a experimental and above. historical data. The report, " and Gamma Doses from an Accidental Critical An excursion is assumed ity," explains this development. * These to occur in a equa vented vessel of unfavorable tions cannot be expected to be as accurate as geometry con taining a solution of detailed calculations based on actual accident 400 g/l of uranium enriched in U-235. The excursion conditions. Comparisons with published in produces an formation indicate they initial burst of 1E+18 fissions in 0.5 second may not be conservative followed for smaller accidents successively at 10-minute intervals by (e.g., 1-2E+17 fissions). 47 bursts However, for accidents of 1.9-E+17 fissions for a total of that are likely to be 1E+19 fissions assumed for safety assessment purposes, in 8 hours. The excursion is they assumed to be appear to be sufficiently conservative. These terminated by evaporation of 100 liters of the solui "ons. 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 criticality accident. RELEASE OF RADIO ACTIVE MATERIAL ARE AS FOLLOWS: 4 C. REGULATORY POSITION a. It should be assumed that all of the noble gas fission products and i. FOLLOWING ARE 25% of the iodine THE PLANT ASSESSMENT AND ASSUMP resulting from TIONS RELATED TO ENERGY the excursion are RELEASE FROM A CRITI released directly to a ventilated CALITY ACCIDENT AND room atmos THE MINIMUM CRITICALITY phere. It should also ACCIDENT TO BE CONSIDERED: be assumed that an aerosol, which is generated from the evapora tion of solution a. When defining the characteristics during the excursion, is of an released directly to the assumed criticality accident in order room atmosphere. The to assess aerosol should be assumed the adequacy of structures, systems, to comprise 0.05%of and com the salt content of ponents provided for the solution that is the .prevention or evaporated. The room mitigation of the consequences volume and air ventila of accidents, tion rate and retention the applicant should evaluate credible time should be con 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 cessing of fissile of radiological decay during materials or into which fissile transit within materials in significant the plant should be evaluated on amounts could be an individual case basis.

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

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 ýinclude the iodine depletion allowance. c. For the first 8 hours, the breathing rate of a person offsite should be assumed to 3 3. ACCEPTABLE ASSUMPTIONS FOR DOSE AND DOSE CON be 3.47E-4m /sec, From 8 to 24 hours following VERSION ARE AS FOLLOWS: the accident, the breathing rate should be as sumed to be 1.75E-4mS/sec. These values were should show that the con developed from the average daily breathing a. The applicant 3 sequences of the prompt gamma and neutron rate (2E+7 cm /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 5 torial additions made so that gamma and beta (!) Prompt Gamma Dose emitting material could be considered.] Under these conditions the rate of energy absorption D = 2.1E-20 N d-2e-s'4d per unit volume is equal to the rate of energy For an infinite where released per unit volume. uniform cloud containing X curies of beta radio the beta dose rate in D I = gamma dose (rcm) activity, per cubic meter, air at the cloud center is N = number of fissions pDo, = 0.475EYx d = distance from source (kin). The surface body dose rate from beta emitters Data presented in The Effects of Nuclear in the infinite cloud can be approximated as being Weapons (Ref. 11, p. 384) may be used to one-half this amount (i.e., pDeI = 0.231 x). 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 air at the cloud center is for the first 8 inches, a factor of 5.0 for the first foot, and a factor of 5.5 for each addi tional foot. YD- = 0.507Eyx

(2) Prompt Neutron Dose From a semi-infinite cloud, the gamma dos6 rate in air is Dn = 7E-20N d'2e-5.2d whefe ¥Do, = 0.25E I X Dn = neutron dose (rem) where N = number of fissions

d = distance from source (kim). Do* = beta dose rate from an infinite cloud (rad/sec) $Most of the gamma radiation is emitted in the actual fission process. Some gamma radiation is produced in various D- = gamma dose rate from an infinite secondary nuclear processes. including decay of fission V •cloud (rad/sec) products. For the purposes of this guide. "prompt* gamma doses should be evaluated including the effects of decay of significant fission products during the first minute of the E = average betaenergyper disintegration excursion. For conditions cited in the example, the equation given includes these considerations. (MeV/dis)

3.34-5 -y = 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 3 fabrication plant 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 I dose calculations: Accidental Nuclear Criticality in a Plutonium Processing and Fuel Fabrication Plant,"' (1) The dose at any distance from I 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 apjpropriate 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 x/Q = I sidered as the dose from gamma radiation at a nua a 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 2 depth of 7 mg/cm . The beta skin dose may be x = the short-term average centerline value estimated by applying of the ground level concentration an energy-dependent (Ci/ms3 attenuation factor (Dd/DB ) to the surface dose ) 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 y the highest radiation dose after the isotope is plume (m). [See Ref. 17, Figure V-I, absorbed into the body. For the purpose of p. 48.] this guide, the following assumptions should be made: az = the vertical standard deviation of the plume (m). [See Ref. 17, Figure V-2, (1) The dose conversion p. 48.1 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 hours, the plume should be assumed to meander and spread uniformly over 6 a 22.50 (2) The effective half-life for the nuclide sector. The resultant equation is is as recommended in ICRP Publication 6 (Ref. I 16) or other x/Q = 2.032 appropriate source. azux g. The potential dose for all significant nuclides should be estimated for the population 6nhe sector may be assumed to shift after 8 hours, if local distribution on a site-related basis. meteorologcal data are available to justify a wind direction. 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 given in b(4). x = distance from point of release to the re based on the parameters ceptor; other variables are given in b(2). 7 (4) The atmospheric diffusion model for D. IMPLEMENTATION releases is based on the informa ground level The purpose of this section is to provide tion in the following table: information to applicants and licensees regard ing the staff's plans for using this regulatory Time guide. Following Accident Atmospheric Conditions Except in those cases in which the applicant proposes an alternative method for complying 0 to 8 hours Pasquill Type F; windspeed with specified portions of the Commission's I m/sec; uniform direction regulations, the method described herein will be used in the evaluation of submittals for 8 to 24 hours Pasquill Type F; windspeed special nuclear material license applications I m/sec; variable direction docketed after December 1, 1977. within a 22.50 sector. If an applicant wishes to use this regulatory ?In some cases site-dependent parameters such as meteo submittals for applications rology. topography, and local geography may dictate the use of guide in developing a more restrictive model to ensure a conservative estimate of docketed on or before December 1, 1977, the Spotential offsite exposures. In such cases, appropriate site pertinent portions of the application will be related meteorology should be developed on an individual case the basis of this guide. basis. evaluated on

3.34-7 REFERENCES

I. W. R. Stratton, "Review of Criticality 12. "Permissible Dose for Internal Incidents," LA-3611, Los Radiation," Alamos Scientifc Publication 2, Report Laboratory (Jan. 1967). of Committee II, 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 United Kingdom Atomic Energy Authority. Energy--1968, 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. Water Boilers," ' NAA-SR-7087, M. Hollander, 1. Perl Atomic man, Table of.Isotopes, International (Feb. 1973). 6th Ed., Lawrence' Radiation Laboratory, Univ. of California, Berkeley, CA (1967). 4. P. Licorchi, R. L. Seale, "A Review of the Experiments Performed to Determine 15. Radiation the Dosimetry, G. J. Hine and G. L. Radiological Consequences of a Criti Brownell, cality Accident," Editors, Academic Press, New Y-CDC-12, Union Carbide York (1956) Corp. (Nov. 1973). 16. Recommendations of ICRP, Publication 5. G. Tuck, "Simplified Methods of Estimating 6, Pergamon Press (1962). the Results of Accidental Solution Excursions," Nuch. Technol., Vol. 23, 17. F. A. Gifford, Jr., p. 177 (1974). "Use of Routine Meteor ological Observations for Estimating Atmos pheric Dispersion," 6. A. R. Olsen, Nuclear Safety, Vol. 2, 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 ANS Trans., Vol. 19, p. 189-91 (1974). Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant 7. T. F. Wimmette et al., "Godiva 2--An Accident Unmoderated for Boiling Water Reactors," U.S. Nuclear Pulse Irradiation Reactor," Regulatory Nucl. Sci. Enz., Commission, Washington, D.C. Vol. 8, p. 691 (1960). (June 1974). 8. W. E. Nyer, G. 0. Bright, R. J. 19. "Radiological McWhorter, Health Handbook," U.S. "Reactor Excursion Behavior," Department International Conference of Health,.- Education and Wel 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. 9. E. D.. Clayton, "Anomalies of Criticality," Meek and B. F. Rider, General Electric Vallecitos Nucl. Technol., Vol. 23, No. 14 (1974). Nuclear 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 United Nuclear, Inc. (June 1968). for Radionuclides Oc curring in Routine Releases from Nuclear Fuel Cycle Facilities," 11. The Effects of Nuclear Weapons, Revised ORNL/NUREG/TM Edition, S. Glasstone, 102, D. C. Kocher, Oak Ridge National Editor, U.S. Dept. Laboratory, of Defense Oak Ridge, Tennessee 37830 (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 Half lifeb, c Nuclide 0-0.5 Hr. 0.5-8 Hr. Total Pb Kr-83m 1.8 h 2.2E+1 1.4E+2 1.6E+2 2.6E-3 0 Kr-85m 4.5 hr 2.1E+1 1.3E+2 1. 5E+2 1.6E-1 2.5E-1 Kr-85 10.7 y 2.2E-4 1.4E-3 1.6E-3 2.2E-3 2.5E-1 Kr-87 76.3 m 1.4E+2 8.5E+2 9.9E+2 7.8E-1 1.3E0 Kr-88 2.8 h 9.1E+1 5.6E+2 6.5E+2 2. OEO 3.5E-1 Kr-89 3.2 m 5. 9E+3 3.6E+4 4.2E+4 1.6E0 1. 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 1.1E-1 Xe-135m 15.6 m 3. 1E+2 1.9E+3 2.2E+3 4.3E-2 9.0E-2 Xe-135 9.1 h 5.0E+1 3. 1E+2 3.6E+2 2.5E-1 3.7E-1 Xe-137 3.8 m 6.9E+3 4.2E+4 4.9E+4 1.6E-1 1.8E0 Xe-138 14.2 m 1.8E+3 1. IE+4 1.3E+4 1. 1E0 6.2E-1

1-131 8.0 d 1. 2E0 7.5E0 8. 7E0 3.18E-1 1.9E-1 1-132 2.3 h 1. 5E+2 9.5E+2 1. 1E+3 2.2E0 5.0E-1 1-133 20.8 h 2.2E+l 1.4E+2 1.6E+2 6. 1E-1 4.1E-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 1.5E0 3.7E-1 aTotal curies are based on cumulative yields for fission energy spectrum using the data in Ret. 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. b Half-llves and average energies are derived using the data in Ret. 21. c. year d * day h s hour m = minute

3.34-9 -I

1 -61-1 ,-IT L011 I 1 -f T I/ I/" L1 I I Y 02 1 I VI /! 10-1 00.1If I /

10-2

ILI

il II 10-3 1 I I1I 1 0.1 1.0 10 Maximum Beta Energy. MeV 1 RATIO OF DEPTH DOSE TO SURFACE DOSE AS A FUNCTION BETA ENERGY SPECTRA for Infinite Plae Source of Infinite Thickness and for Allowed Spectra 1 DeOloped from Consideratson, Presented in Reference 1S. Chapter 16

FIGURE I

3.34-10 Building Wake Dispersion Correction Factor

00

ix n

30

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i 10-2

10 ""-24 hours

M.

1o2 1o3 1l'

Distance from Structure Imetars) FIGURE 3A (Ref. 18)

3.34-12 FisUmE ham UunR .1n8u) FIGURE 35 (Rdl. 18)

3.34-13 UNITEDSTATES NUCLEAR REGULATORYCOMMISSION WASHINGTON, b. C. 20555 POSTAGE AND FEESPAID OFFICIAL BUSINESS UNITU STAlPS PNCLLA* PENALTY FOR PRIVATEUSE, $300 Ri•GAAoIT COMMI5SION