Regulatory Guide 3.35: Difference between revisions

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{{Adams
{{Adams
| number = ML12220A062
| number = ML003739504
| issue date = 07/31/1979
| issue date = 07/31/1979
| title = Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Plutonium Processing and Fuel Fabrication Plant
| title = Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Plutonium Processing and Fuel Fabrication Plant. (Withdrawn 1/15/98)
| author name =  
| author name =  
| author affiliation = NRC/RES, NRC/OSD
| author affiliation = NRC/RES
| addressee name =  
| addressee name =  
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| license number =  
| license number =  
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| document report number = RG-3.035, Rev. 1
| document report number = RG-3.35, Rev.1
| document type = Regulatory Guide
| document type = Regulatory Guide
| page count = 20
| page count = 20
}}
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{{#Wiki_filter:U.S. NUCLEAR REGULATORY
{{#Wiki_filter:}}
COMMISSION
Revision 1 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==
Section 70.22, "Contents of Applications," of 10 CFR Part 70, "Domestic Licensing of Special Nuclear Materials," requires, that each appli-cation for a license to possess and use special nuclear material in a plutonium processing and fuel fabrication plant contain a description and safety assessment of the design bases of the principal structures, systems, and components of the plant. Section 70.23(a)(3)
states that applications will be approved if the Commission determines that, among other factors, the applicant's proposed equipment and facilities are adequate to protect health and minimize danger to life and property, and Sec-tion 70.23(b) states that the Commission will approve construction of the principal struc-tures, systems, and components of the plant when the Commission has determined that the design bases of the principal structures, sys-tems, and components and the quality assurance program provide reasonable assurance of protection against the consequences of potential accidents.
 
In plutonium processing and fuel fabrication plants, a criticality accident is one of the postulated accidents used to evaluate the ade-quacy of an applicant's proposed activities with respect to the public health and safety. This guide describes methods used by the NRC staff in the analysis of such accidents.
 
These methods result from review and action on a number of specific cases and, as such, reflect the lates~t general NRC-approved approaches to the problem. If an applicant desires to employ new information that may be developed in the future or to use an alternative method, NRC*Lines indicate substantive changes from previous issue.will review the proposal and approve its use, if found acceptable.
 
==B. DISCUSSION==
In the process of reviewing applications for permits and licenses authorizing the construc-tion or operation of plutonium processing and fuel fabrication plants, the NRC staff has developed a number of appropriately conser-vative assumptions that are used by the staff to evaluate an estimate of the radiological consequences of various postulated accidents.
 
These assumptions are based on previous accident experience, engineering judgment, and on the analysis of applicable experimental results from safety research programs.
 
This guide lists assumptions used by the staff to evaluate the magnitude and radiological conse-quences of a criticality accident in a plutonium processing and fuel fabrication plant.A criticality accident is an accident resulting in the uncontrolled release of energy from an assemblage of fissile material.
 
The cir-cumstances of a criticality accident are difficult to predict. However, the most serious criticality accident would be expected to occur when the reactivity (the extent of the deviation from criticality of a nuclear chain reacting medium) could increase most rapidly and without control in the fissile accumulation of the largest credible mass. In plutonium pro-cessing and fuel fabrication plants where con-ditions that might lead to criticality are carefully avoided because of the potential for adverse physical and radiological effects, such an accident is extremely uncommon.
 
However, experience with these and related facilities has demonstrated that criticality accidents may occu
 
====r. 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 plants, such an accident might be initiated by (1) the inadvertent transfer or leakage of a solution of fissile material from a geometrically safe containing vessel into an area or vessel not so designed, (2) introduction of excess fissile material solution to a vessel, (3) intro-duction of excess fissile material to a solution, (4) overconcentration of a solution, (5) preci-pitation of fissile solids from a solution and their retention in a vessel, (6) introduction of neutron moderators or reflectors (e.g., entrance of water to a higly under-moderated system), (7) deformation of or failure to maintain safe storage arrays, or (8) similar actions which can lead to increases in the reactivity of fissile systems. Some acceptable means for minimizing the likelihood of such accidents are described in Regulatory Guide 3.4, "Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.
 
"1 I. CRITICALITY
ACCIDENT EXPERIENCE
IN RELATION TO THE ESTIMATION
OF THE MOST SEVERE ACCIDENT Stratton (Ref. 1) has reviewed in detail 34 occasions prior to 1966 when the power level of a fissile system increased without control as a result of unplanned or unexpected changes in its reactivity.
 
Although only six of these incidents occurred in processing operations, and the remainder occurred mostly in facilities for obtaining criticality data or in experimental reactors, the information obtained and its correlation with the characteristics of each system have been of considerable value for use in estimating the consequences of accidental criticality in process systems. The incidents occurred in aqueous solutions of uranium or plutonium
(10), in metallic uranium or plutonium in air (9), in inhomogeneous water-moderated systems (9), and in miscellaneous solid uranium systems (6). Five occurred in plutonium systems, including reactors and criticality studies, of which three were in solutions.
 
The estimated total number of fissions per incident ranged from 1E+15 2 to 1E+20 with a median of about 2E+17. More recently, another incident in a plutonium processing facility at Windscale (U.K.) was described in which a total yield of about 1E+15 fissions apparently occurred (Ref. 2). In ten cases, the supercriticality was halted by an automatic control device. In the remainder, the shutdown was effected as a consequence of the fission energy release which resulted in thermal expansion, density reduction from the formation of very small bubbles, mixing of light'Copies may be obtained from the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention;
Director, Division of Document Control.2 1E÷15 = 1 x 1015. This notational form will be used in this guide.and dense layers, loss of water moderator by boiling, or expulsion of part of the mass.Generally, the criticality incidents were characterized by an initial burst or spike in the curve of fission rate versus time followed by a rapid but incomplete decay of the fission rate as the shutoff mechanism was initiated.
 
As more than one shutdown mechanism may affect the reactivity of the system and the effect of a particular mechanism may be counteracted, the initial burst was frequently succeeded by a plateau period of varying length. This plateau was characterized by a lesser and declining fission rate and finally by a further dropoff as shutdown was completed.
 
The magnitude of the initial burst was directly related to the rate of increase of reactivity and its magnitude above the just-critical value but was inversely related to the background neutron flux, which is much greater for plutonium than for uranium systems.Those systems consisting only of solid fissile, reflector, or moderator materials exhibited little or no plateau period, whereas solution systems had well developed plateaus.For solution systems, the energy release during the plateau period, because of its dura-tion, provided the major portion of total energy released.
 
For purposes of the planning neces-sary to deal adequately with criticality incidents in experimental and production-type nuclear facilities, Woodcock (Ref. 3) made use of these data to estimate possible fission yields from excursions in various types of systems.For example, spike yields of 1E+17 and 1E+18 and total yields of 3E+18 and 3E+19 fissions were suggested for criticality accidents occurring in solution systems of 100 gallons or less and more than 100 gallons, respectively.
 
Little or no mechanical damage was predicted at these levels.
 
===2. METHODS DEVELOPED ===
FOR PREDICTING
THE MAGNITUDE OF CRITICALITY
ACCIDENTS The nuclear excursion behavior of solu-tions of enriched uranium has been studied extensively both theoretically and experi-mentally.
 
A summary by Dunenfeld and Stitt (Ref. 4) of the kinetic experiments on water boilers, using uranyl sulfate solutions, describes the development of a kinetic model that was confirmed by experiment.
 
This model defines the effects of thermal expansion and radiolytic gas formation as power-limiting and shutdown mechanisms.
 
The results of a series of criticality excur-sion experiments resulting from the introduc-tion of uranyl nitrate solutions to vertical cylindrical tanks at varying rates are sum-marized by Ldcorchd and Seale (Ref. 5). This report confirms the applicability of the kinetics model for solutions, provides correlations of peak power with reactivity addition rate, notes-J 3.35-2 the importance of a strong neutron source in limiting peak power, and indicates the nature of the plateau following the peak.Many operations with fissile materials in a plutonium processing plant may be conducted with aqueous (or organic solvent) solutions of fissile materials.
 
Consequently, well-founded methods for the prediction of total fissions and maximum fission rate for accidents that might occur in solutions (in process or other vessels)by the addition of fissile materials should be of considerable value in evaluating the effects of possible plutonium processing plant criticality accidents.
 
From the results of excursion studies and from accident data, Tuck (Ref. 6)has developed methods for estimating
(1) the maximum number of fissions in a 5-second interval (the first spike), (2) the total number of fissions, and (3) the maximum specific fis-sion rate in vertical cylindrical vessels, 28 to 152 cm in diameter and separated by >30 cm from a bottom reflecting surface, resulting from the addition of up to 500 g/1l solutions of Pu-239 or U-235 to the vessel at rates of 0.7 to 7.5 gal/min. Tuck also gives a method for estimating the power level from which the steam-generated pressure may be calculated and indicates that use of the formulas for tanks>152 cm in diameter is possible with a loss in accuracy.Methods for estimating the number of fis-sions in the initial burst and the total number of fissions, derived from the work reported by L6corchi and Seale (Ref. 5), have also been developed by Olsen and others (Ref. 7). These were evaluated by application to ten actual accidents that have occurred in solutions and were shown to give conservative estimates in all cases except one.Fission yields for criticality accidents occurring in solutions and some heterogeneous systems, e.g., aqueous/fixed geometry, can be estimated with reasonable accuracy using existing methods. However, methods for esti-mating possible fission yield from .other types of heterogeneous systems, e.g., aqueous/powder, are less reliable because of the uncertainties involved in predicting the reactivity rate. The uncertainty of geometry and moderation results in a broad range of possible yields.Woodcock (Ref. 3) estimated that in solid plutonium systems, solid uranium systems, and heterogeneous liquid/powder systems (fissile material not specified)
total fission yields (sub-stantially occurring within the spike) of 1E+18, 3E+19, and 3E+20, respectively, could be predicted.
 
Mechanical damage varied from slight to extensive.
 
Heterogeneous systems consisting of metals or solids in water were estimated to achieve a possible magnitude of 1E+19 following an initial burst of 3E+18 fissions.
 
The possibility of a burst of 3E+22 fissions resulting in a serious explosion could be conceived for large storage arrays where prompt criticality was exceeded, e.g., by collapse of shelving.
 
It is recognized that in such arrays, where reactivity is more likely to be increased by the successive additions of small increments of materials, only a delayed critical condition with maximum yields of 1E+19 fissions is likely. These estimates could aid in the analysis of situations in plant systems.However, they should not be taken as absolute values for criticality assumptions for the purpose of this guide.For systems other than solution systems, the estimation of the peak fission rate and the total number of fissions accompanying an acci-dental nuclear criticality may be estimated with the aid of information derived from accident experience and from the SPERT-l reactor tran-sient tests with light- and heavy-water moderated uranium-alumium and U0 2-stainless steel clad fuels (Ref. 8). Oxide core tests in the latter group provide some information on energy release mechanisms that may be effective, for example, in fabricated fuel element storage in a mixed oxide fuel fabrica-tion plant. Review of unusal process struc-tures, systems, and components for the possibility of. accidental criticality should also consider recognized anomalous situations in which the possibility of accidental nuclear cri-ticality may be conceived (Ref. 9).The application of the double-contingency principle 3 to fissile material processing opera-tions has been successful in reducing the probability of accidental criticality to a low value. As a consequence, the scenarios required to arrive at accidental criticality involve the assumption of multiple breakdowns in the nuclear criticality safety controls.
 
It has therefore been a practice to simply and conservatively as'sume an accidental criticality of a magnitude equal to, or some multiple of, the historical maximum for all criticality acci-dents outside reactors without using any scenario clearly defined by the specific opera-tions being evaluated.
 
In the absence of sufficient guidance, there has been wide vari-ation in the credibility of the postulated magnitude of the occurrence (particularly the size of the initial burst), the amount of energy and radioactivity assumed to be released, and the magnitude of the calculated consequences.
 
It is the staff's judgment that the evalua-tion of the criticality accident should assume the simultaneous breakdown of at least two independent controls throughout all elements of the operation.
 
Each control should be such that its circumvention is of very low probability.
 
Experience has shown that the simultaneous
3The double-contingency principle is defined in ANSI N16. 1-1975, "Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," which is endorsed by Regulatory Guide 3.4.3.35-3 failure of two independent controls is very unlikely if the controls are derived, applied, and maintained with a high level of quality assurance.
 
However, if controls highly dependent on human actions are involved, this approach will call for some variation in the assumed number of control failures.
 
The criticality accidents so conceived should then be analyzed to determine the most severe within the framework of assumed control failures, using realistic values of such variables as the fissile inventory, vessel sizes, and pump transfer rates.
 
===3. RADIOLOGICAL ===
CONSEQUENCES
OF ACCIDENTAL
CRITI-CALITY Past practice has been to evaluate the radiological consequences to individuals of postulated accidental criticality in plutonium processing and fuel fabrication plants in terms of a fraction of the guideline values in 10 CFR Part 100, "Reactor Site Criteria." The consequences of a criticality accident may be limited by containment, shielding, isolation distance, or evacuation of adjacent occupied areas subsequent to detection of the accident.
 
If the impact of a criticality accident is to be limited through evacuation of adjacent occupied areas, there should be prior formal arrangements with individual occupants and local authorities sufficient to ensure that such movements can be effected in the time allowed.The equations provided for estimating doses from prompt gamma and neutron radiation were developed using experimental and historical data. The report, "Promp Neutron and Gamma Doses from an Accidental Criticality," explains this development.*
These equations cannot be expected to be as accurate as detailed calculations based on actual accident conditions.
 
Comparisons with published information indicate they may not be conservative for smaller accidents
.(e.g. , 1-2E+17 fissions).
However, for accidents that are likely to be assumed for safety assessment purposes, they appear to be sufficiently conservative.
 
These equations are included in the guide to provide a simplified method for estimatinK
prompt gamma and neutron radiation doses from a potential criticality accident.C. REGULATORY
POSITION I. FOLLOWING
ARE THE PLANT ASSESSMENT
AND ASSUMP-TIONS RELATED TO ENERGY RELEASE FROM A CRITI-CALITY ACCIDENT AND THE MINIMUM CRITICALITY
ACCIDENT TO BE CONSIDERED:
a. When defining the characteristics of an assumed criticality accident in order to assess*A copy of Charles A. Willis' report, "Prompt Neutron and Gamma Doses, from an Accidental Criticality," is available for inspection at the NRC Public Document Room, 1717 H Street NW., Washington, D.C.the adequacy of structures, systems, and components provided for the prevention or mitigation of the consequences of accidents, the applicant should evaluate credible criticality accidents in all those elements of the plant provided for the storage, handling, or processing of fissile materials or into which fissile materials in significant amounts could be introduced.
 
To determine the circumstances of the criticality accidents, controls judged equivalent to at least two highly reliable, independent criticality controls should be assumed to be circumvented.
 
The magnitude of the possible accidents should then be assessed, on an individual case basis, to estimate the extent and nature of possible effects and to provide source terms for dose calculations.
 
The most severe accident should then be selected for the assessment of the adequacy of the plant. In order to determine the source terms for release of plutonium, the powder mixture should be the maximum weight percent pluto-nium to uranium compound to be used in a mixed oxide fuel fabrication plant.Calculation of the radioactivity of fis-sion products may be accomplished by computer code RIBD (Ref. 10). An equivalent calculation may be substituted, if justified on an individual case basis.b. If the results of the preceding evalu-ation indicate that no possible criticality accident exceeds in severity the criticality accident postulated in this section, then the conditions of the following example may be assumed for the purpose of assessing the adequacy of the facility.
 
A less conservative set of conditions may be used if they are shown to be applicable by the specific analyses conducted in accordance with paragraph C.l.a above.An excursion that produces an initial burst of 1E+18 fissions in 0.5 seconds followed successively at 10-minute intervals by 47 bursts of 1.9E+17 fissions for a total of 1E+19 fissions in 8 hours is assumed to occur.The excursion is assumed to be terminated by evaporation of 100 liters of the solution.
 
===2. ASSUMPTIONS ===
RELATED TO THE RELEASE OF RADIO-ACTIVE MATERIAL ARE AS FOLLOWS: 4 a. It should be assumed that all of the noble gas fission products and 25% of the iodine radionuclides are released directly to a ventilated room whose construction is typical of the plant's Class I structures.
 
If the accident is assumed to occur in a solution, it should also be assumed that an aerosol, which is generated from the evaporation of solution during the excursion, is released directly to the room atmosphere.
 
The aerosol should be assumed to 4Certain assumptions for release of radioactive material, dose conversion, and atmospheric diffusion reflect the staff's position indicated in Regulatory Guide 1.3 (Ref. 20).3.35-4 comprise 0.05% of the salt content of the solution that is evaporated.
 
The room volume and ventilation rate and retention time should be considered on an individual case basis.b. The effects of radiological decay during transit within the plant and in the plant exhaust system should be taken into account on an individual case basis.c. The reduction in the amount of radio-active material available for release to the environment through the plant stack as a result of the normal operation of filtration systems in the plant exhaust systems may be taken into account, but the amount of reduc-tion in the concentration of radioactive mate-rials should be evaluated on an individual case basis.d. Table 1 lists the radioactivity of sig-nificant nuclides released, but it does not include the iodine depletion allowance.
 
*
 
===3. ACCEPTABLE ===
ASSUMPTIONS
FOR DOSE AND DOSE CON-VERSION ARE AS FOLLOWS: a. The applicant should show that the con-sequences of the prompt gamma and neutron dose are sufficiently mitigated to allow occupancy of areas necessary to maintain the plant in a safe condition following the accident.The applicant should estimate the prompt gamma and neutron doses that could be received at the closest site boundary and nearest residence.
 
The following semi-empirical equations may be used for these calculations.
 
Because detailed evaluations will be dependent on the site and plant design, different methods may be substituted on an individual case basis.Potential total dose attenuation due to shielding and dose exposures should be evaluated on an individual case basis.(I) Prompt 5 Gamma Dose D = 2.IE-20 Nd-2 e-3.4d w where first foot, and a factor of 5.5 for each addi-tional foot.(2) Prompt Neutron Dose Dn = 7E-20 Nd" 2 e-5.2d where Dn = neutron dose (rem)N = number of fissions d = distance from source (kin)For concrete, the dose should be reduced by a factor of 2.3 for the first 8 inches, 4.6 for the first foot, and a factor of 20 for each additional foot.b. No correction should be made for deple-tion from the effluent plume of radioactive iodine due to deposition on the ground or for the radiological decay of iodine in transit.c. For the first 8 hours, the breathing rate of a person off site should be assumed to be 3.47E-4 mS/sec. From 8 to 24 hours follow-ing the accident, the breathing rate should be assumed to be 1.75E-4 m 3/sec. These values were developed from the average daily breath-ing rate (2E + 7 cm 3/day) assumed in the report of ICRP Committee
11-1959 (Ref. 12).d. External whole body doses should be calculated using "Infinite Cloud" assumptions, i.e., the dimensions of the cloud are assumed to be large compared to the distance that the gamma rays and beta particles travel. "Such a cloud would be considered an infinite cloud for a receptor at the center because any additional (gamma and] beta emitting material beyond the cloud dimensions would not alter the flux of[gamma rays and] beta particles to the receptor." [See Meteorology and Atomic Energy--1968 (Ref. 13), Section 7.4.1.1;editorial additions made so that gamma and beta emitting material could be considered.]
Under these conditions, the rate of energy absorption per unit volume is equal to the rate of energy released per unit volume. For an infinite uniform cloud containing X curies of beta radioactivity per cubic meter, the beta dose rate in air at the cloud center is D- = 0.457EPX The surface body dose rate from beta emitters in the infinite cloud can be approximated as being one-half this amount (i.e., pDoo = 0.23EX).For gamma emitting material, the dose rate in air at the cloud center is Do, = o.5o07E X Y D = gamma dose (rein)¥N = number of fissions d = distance from source (kin)Data presented in The Effects of Nuclear Weapons (Ref. 11, p. 384) may be used to 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 Syost of the gamma radiation is emitted in the actual fission process. Some gamma radiation is produced in various second-ary nuclear processes, including decay of fission 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 excursion.
 
For conditions cited in the example, the equation given includes these considerations.
 
3.35-5 From a semi-infinite cloud, the gamma dose rate in air is= o.25EYx where I D-= beta dose rate from an infinite cloud (rad/sec)Do, = gamma dose rate from an infinite¥ cloud (rad/sec)E = average beta energy per disintegration (MeV/dis)EY = average gamma energy per disintegration
¥ (MeV/dis)X = concentration of beta or gamma emitting isotope in the cloud (Ci/m 3)e. The following specific assumptions are acceptable with respect to the radioactive cloud dose calculations:
(1) The dose at any distance from the plant should be calculated based on the maxi-mum concentration time integral (in the course of the accident)
in the plume at that distance, taking into account specific meteorological, topographical, and other characteristics that may affect the maximum plume concentration.
 
These site-related characteristics should be evaluated on an individual case basis. In the case of beta radiation, the receptor is assumed to be exposed to an infinite cloud at the maximum ground level concentration at that distance from the plant. In the case of gamma radiation, the receptor is assumed to be exposed to only one-half the cloud owing to the presence of the ground. The maximum cloud concentration should always be assumed to be at ground level.(2) The appropriate average beta and gamma energies emitted per disintegration may be derived from the Table of Isotopes (Ref. 14)or other appropriate sources, e.g. , Ref. 23.(3) The whole body dose should be considered as the dose from gamma radiation at a depth of 5 cm and the genetic dose at a depth of 1 cm. The skin dose should be the sum of the surface, gamma dose and the beta dose at a depth of 7 mg/cm 2.The beta skin dose may be estimated by applying an energy-dependent attenuation factor (Dd/DB) to the surface dose according to a method developed by Loevinger, Japha, and Brownell (Ref. 15).(See Figure 1.)f. The "critical organ" dose from the in-haled radioactive materials should be estimated.
 
The "critical organ" is that organ that receives the highest radiation dose after the isotope is absorbed into the body. For the purpose of this guide, the following assumptions should be made: (1) The radionuclide dose conversion factors are as recommended by the report of Committee
11, ICRP (Ref. 12) or other appro-priate source.(2) The effective half-life for the nu-clide is as recommended in ICRP Publication
6 (Ref. 16) or other appropriate source.(3) The plutonium and other actinide nuclide clearance half time, or fraction of nu-clide clearing the organ, is as recommended by the ICRP task group on lung dynamics (Ref. 17). A computer code, DACRIN (Ref. 18), is available for this model. Task group lung model (TGLM) clearance parameters are presented in Table 2; the model is shown schematically in Figure 2.g. The potential dose exposure for all sig-nificant nuclides should be estimated for the population distribution on a site-related basis.
 
===4. ACCEPTABLE ===
ASSUMPTIONS
FOR ATMOSPHERIC
DIFFU-SION ARE AS FOLLOWS: a. Elevated releases should be considered to be at a height equal to not more than the actual stack height.6  Certain site-dependent conditions may exist, such as surrounding elevated topography or nearby structures, that will have the effect of reducing the actual stack height. The degree of stack height reduction should be evaluated on an individual case basis.Also, special meteorological and geo-graphical conditions may exist that can con-tribute to greater ground level concentrations in the immediate neighborhood of a stack. For example, fumigation should always be assumed to occur; however, the length of time that a fumigation condition
.exists is strongly dependent on geographical and seasonal factors and should be evaluated on a case-by-case basis.' (See Figure 3 for elevated releases under fumigation conditions.)
b. For plants with stacks, the atmospheric diffusion model should be as follows: Scredit for an elevated release should be given only if the point of release is (1) more than two and one-half times the height of any structure close enough to affect the dispersion of the plume or (2) located far enough from any structure that could have an effect on the dispersion of the plume. For those plants without stacks, the atmospheric diffusion factors assuming ground level releases, as shown in Regulatory Position 4.c, should be used.7 For sites located more than 2 miles from large bodies of water, such as oceans or one of the Great Lakes, a fumigation condition should be assumed to exist at the time of the accident and continue one-half hour. For sites located less than 2 miles from large bodies of water, a fumigation condition should be assumed to exist at the time of the accident and continue for 4 hours.I 3.35-6
(1) The basic equation for atmospheric diffusion from an elevated release is exp(-he 2/2Cz 2 )X/Q =iiua a yz where x = the short-term average centerline value of the ground level concentration (Ci/m 3)Q = rate of material release (Ci/sec)u = windspeed (m/sec)a = the horizontal standard deviation of the Y plume (m). [See Ref. 19, Figure V-l, p. 48.]a = the vertical standard deviation of the z plume (m). [See Ref. .19, Figure V-2, p. 48.]h = effective height of release (m)8 e (2) For time periods of greater than 8 hours, the plume from an elevated release should be assumed to meander and spread uniformly over a 22.50 sector.9 The resultant equation is 8 to 24 hours suming various stack heights] windspeed
1 m/ sec;uniform direction.
 
See Figure 5 for Envelope of-Pasquill diffusion categories;
windspeed
1 m/sec; variable direction within a 22.50 sector.x/Q =2.032 exp(-h e2/2 z2)e UX y ux z where x = distance from the release point (m);other variables are as given in b(l).(3) The atmospheric diffusion model'0 for an elevated release as a function of the distance from the plant is based on the infor-mation in the following table.c. If no onsite meteorological data are available for facilities exhausted wihout stacks, or with stacks that do not meet the elevated release criteria, the atmospheric diffusion model should be as follows: (1) The 0-to-8 hour ground level re-lease concentrations may be reduced by a factor ranging from one to a maximum of three (see Figure 6) for additional dispersion produced by the turbulent wake of a major building in calculating nearby potential expo-sures. The volumetric building wake correction factor, as defined in Section 3.3.5.2 of Meteorology and Atomic Energy--1968 (Ref. 13), should be used in the 0-to-8 hour period only; it is used with a shape factor of one-half and the minimum cross-sectional area of a major building only.(2) The basic equation for atmospheric diffusion from a ground level point source is x/Q= 1 nuraa yz where X = the short-term average centerline value of the ground level concentration (Ci/m 3)Q = rate of material release (Ci/sec)u = windspeed (m/sec)a = the horizontal standard deviation of the y plume (m) [see Ref. 19, Figure V-i, p. 48]a = the vertical standard deviation of the z plume (m) [see Ref. 19, Figure V-2, p. 481 (3) For time periods of greater than 8 hours, the plume should be assumed to meander and spread uniformly over a 22.50 sector.9 The resultant equation is 2.032 x/Q= a ux z where X = distance from point of release to the receptor;
other variables are as given in c(2).Time Following Accident Atmospheric Conditions
0 to 8 hours See Figure 4 for Envelope of Pasquill diffusion categories
[based on Figure A7, Meteorology and Atomic Energy--1968 (Ref. 13), as-8h =h -h , where h is the height of the release above plant grads, SanA ht is tde maximum terrain height, above plant grade, between the point of release and the point at which the calculation is made, he should not be allowed to exceed hs.gThe sector may be assumed to shift after 8 hours if local meteorological data are available to justify a wind direction change. This should be considered on an individual case basis.'l°n some cases, site-dependent parameters such as meteor-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-related meteorology should be developed on an indivdual case basis.3.35-7
(4) The atmospheric diffusion model'0 for ground level releases is based on the infor-mation in the following table.Time Following Accident 0 to 8 hours 8 to 24 hours Atmospheric Conditions Pasquill Type F, windspeed 1 m/sec, uniform direction Pasquill Type F, windspeed 1 m/sec, variable direction within a 22.50 sector.
 
==D. IMPLEMENTATION==
The purpose of this section is to provide information to applicants and licensees regard-ing the staff's plans for using this regulatory guide.Except in those cases in which the applicant proposes an alternative method for complying with specified portions of the Commission's regulations, the method described herein will be used in the evaluation of submittals for special nuclear material license applications docketed after December 1, 1977.If an applicant wishes to use this regulatory
*guide in developing submittals for applications docketed on or before December 1, 1977, the pertinent portions of the application will be evaluated on the basis of this guide.(5) Figures 7A and 7B give the ground level release atmospheric diffusion factors based on the parameters given in c(4).I 3.35-8 REFERENCES
1. W. R. Stratton, "Review of Criticality Incidents," LA-3611, Los Alamos Scientific Laboratory (Jan. 1967).2. T. G. Hughes, "Criticality Incident at Windscale," Nuclear Engineering Inter-national, Vol. 17, No. 191, pp. 95-7 (Feb. 1972).3. E. R. Woodcock, "Potential Magnitude of Criticality Accidents," AHSP(RP) R-14, United Kingdom Atomic Energy Authority.
 
4. M. S. Dunenfeld, R. K. Stitt, "Summary Review of the Kinetics Experiments on Water Boilers." NAA-SR-7087, Atomic International (Feb. 1973).5. P. Lgcorch6, R. L. Seale, "A Review of the Experiments Performed to Determine the Radiological Consequences of a Criticality Accident, " Y-CDC-12, Union Carbide Corp.(Nov. 1973).6. G. Tuck, "Simplified Methods of Estimating the Results of Accidental Solution Excur-sions," Nucl. Technol., Vol. 23, p. 177 (1974).7. A. R. Olsen, R. L. Hooper, V. 0. Uotinen, C. L. Brown, "Empirical Model to Estimate Energy Release from Accidental Criticality," ANS Trans., Vol. 19, pp. 189-91 (1974).8. W. E. Nyer, G. 0. Bright, R. J. McWhorter,"Reactor Excursion Behavior," International Conference on the Peaceful Uses of Atomic Energy, paper 283, Geneva (1966).9. E. D. Clayton, "Anomalies of Criticality," Nucl. Technoj., Vol. 23, No. 14 (1974).10. R. 0. Gumprecht, "Mathematical Basis of Computer Code RIBD," DUN-4136, Douglas United Nuclear, Inc. (June 1968).11. The Effects of Nuclear Weapons, Revised Edition, Samuel Glasstone, Editor, U.S.Dept. of Defense (Feb. 1964).12. "Permissible Dose for Internal Radiation," Publication
2, Report of Committee II, International Committee on Radiological Protection (ICRP), Pergamon Press (1959).13. Meteorology and Atomic Energy-- 1968, D. H. Slade, Editor, U.S. Atomic Energy Commission (July 1968).14. C. M. Lederer, J. M. Hollander, I. Perl-man, Table of Isotopes, 6th Edition, Lawrence Radiation Laboratory, Univ. of California, Berkeley, California
(1967).15. Radiation Dosimetry, G. J. Hine and G. L.Brownell, Editors, Academic Press, New York (1956).16. Recommendations of ICRP, Publication
6, Pergamon Press (1962).17. "The Metabolism of Compounds of Plutonium and Other Actinides," a report prepared by a Task Group of Committee II, ICRP, Pergamon Press (May 1972).18. J. R. Houston, D. L. Strenge, and E. C.Watson, "DACRIN--A
Computer Program for Calculating Ocean Dose from Acute or Chronic Radionuclide Inhalation," BNWL-B-389(UC-4), Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, Washington, (Dec. 1974).19. F. A. Gifford, Jr., "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," Nuclear Safety, Vol. 2, No. 4, p. 48 (June 1961).20. Regulatory Guide 1.3, "Assumptions Used for Evaluating the Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors," U. S. Nuclear Regulatory Commission, Washington, D. C.21. J. M. Selby, et al., "Considerations in the Assessment of the Consequences of Efflu-ents from Mixed Oxide Fuel Fabrication Plants," BNWL-1697, Rev. 1 (UC-41), Pacific Northwest Laboratories, Richland, Washington (June 1975)..22. "Compilations of Fission Product Yields," NEDO-12154-1, M. E. Meek and B. F.Rider, General Electric Vallecitos Nuclear Center, TIC, P.O. Box 62, Oak Ridge, Tennessee
37830 (January 1974).23. "Nuclear Decay Data for Radionuclides Occurring in Routine Releases from Nuclear Fuel Cycle Facilities," ORNL/ NUREG/TM-102, D.C. Kocher, Oak Ridge National Laboratory, Oak Ridge, Tennessee
37380 (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 Nuclide Half-life b a Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-131m Xe-133m Xe-133 Xe- 135m Xe- 135 Xe- 137 Xe- 138 1-131 1-132 1-133 1-134 ,1-135 1.8 4.5 10.7 76.3 2.8 3.2 h h y m h m 0-0.5 Hr.1. 5E+1 9.9E0 1. 2E-4 6. OE+1 3'. 2E+1 1.8E+3 0.5-8 Hr. Total 11.9 d 2.0 d 5.2 d 15.6 m 9.1 h 3.8 m 14.2 m 8.0 d 2.3 h 20.8 h 52.6 m 6.6 h 1.4E-2 3.1E-1 3.8E0 4.6E+2 5. 7E+ 1 6.9E+3 1. 5E+3 1. 5E0 1.7E+2 2.2E+l 6.OE+2 6.3E+l 9.5E+1 6. 1E+I 7.2E-4 3.7E+2 2.0E+2 1. 1E+4 8.6E-2 1. 9E0 2.3E+1 2.8E+3 3.5E+2 4.2E+4 9.5E+3 9.5E0 1.OE+3 1.4E+2 3.7E+3 3.9E+2 1. 1E+2 7. IE+1 8. 1E-4 4.3E+2 2.3E+2 1. 3E+4 1.OE-1 2.2E0 2. 7E+I 3.3E+3 4. IE+2 4.9E+4 1. IE+4 1. 1E+I 1.2E+3 1.6E+2 4.3E+3 4.5E+2 5.9E-4 2.7E-5 5.8E-5 1.8E-2 4.3E-7 2.41E-5 C Y 2.6E-3 1.6E-1 2.2E-3 7.8E-1 2. OEO 1.6E0 2.OE-2 4. 1E-2 4.6E-2 4.3E-1 2.5E-1 1.6E-1 1. lEO 3.8E-1 2.2E0 6. IE-1 2.6E0 1.5E0 C 0 2.5E-1 2.5E-1 1. 3E0 3.5E-1 1. 3E0 1.4E-1 1.9E-1 1.1E-1 9.OE-2 3.7E-1 1. 8EO 6.2E-1 1.9E-1 5.OE-1 4. 1E-1 6.1E-1 3.7E-1 Pu-2 3 8 d Pu-239 Pu-240 Pu-241 Pu-242 Am-241 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
NP a b TB c d CLASS Dbc d fd 0.01 0.5 0.01 0.5 0.01 0.95 0.2 0.05 CLASS Wc Td 0.01 0.4 0.01 0.2 CLASS yC d k fd k 0.1 0.01 0.9 0.5 0.5 0.4 0.01 0.2 k 0.01 0.99 0.01 0.99 0.05 0.4 0.4 0.15 0.9 P e f g h L i 0.5 n.a.e n.a.0.5 0.5 0.8 n.a.n.a.0.2 1.0 50 1.0 50 50 50 0.15 500 0.4 0.4 1.0 500 0.05 500 1.0 1000 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 10 A f L0.05 I._ : A...../ -10-2 S I 0.0. 012 10-3 L S0.FIGUR I 0.1 1 3. 10.a. .n.u. Bea E ery e RAIOO IET DOS TO SUFC DOEAI!UCTO
EAEERYSETA
fo ni it I Pln oreo InIntThcesadfoAlwdSptr
...1.De0e2oped frmCosdrain Prsne inRfrnc 5 hatr1: : : FIGURE 1: :: : : 3:.3 5-12; #
LM, CLMF L SCHEMATIC
DIAGRAM DEVELOPED
FROM ICRP TASK GROUP LUNG MODEL (Refil17 FIGURE 2 I.3.35-13
10-2 i ..... -..-:-.' .
........ .ELEVATED RELEASE
* ATMOSPHERIC
DISPERSION
FACTORS
 
====i. FOR FUMIGATION ====
CONDITIONS
T---ATMOSPHERIC
CONDITIONS-
PASQUILL TYPE F WINDSPEED
1 METERISEC-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-E0h =125 meters*h = 150 meters 10-6 10-7 1 1 1 1 idl I I L -I I sall 102 103 104 Distance from Release Point (meters)FIGURE 4 (Ref. 20)3.35-15 ELEVATED RELEASE-... ATMOSPHERIC
DIFFUSION
FACTORS .-.... 8--24 HOUR RELEASE TIME-, L.- .-.-r".-,-- --f- -V --7 -- .. ., , .. ...----,- ---I I I i I I NýDistance from Release Point (meters)FIGURE 5 (Ref. 20)3.35-16
-w 3 2.5 0.5A = 500 meters 2 0.5 2 0.5A = 1000 meters 2 0.5A = 1500 meters 2 , 0.5A = 2000 meters 2.1.5 CoC 0.5 0-102 103 Distance from Structure (meters)FIGURE 6 (Ref. 20)104 I i i i i.i i i -I i i 10-10-: : *-.GROUND
LEVEL RELEASE-ATMOSPHERIC
DIFFUSION
FACTORS FOR VARIOUS TIMES FOLLOWING
ACCIDENT w+- howun..-- + +]* I ............_ I ." --4 --L -... ...... z,- -- --l f ..-24 houri, i.4-"-- I 5 .--102 103 Disunce from Structure (mutre)FIGURE 7A (Ref. 20)105 K 3.35-18 fo'Distman from 8tructure (metesr)FIGURE 7B (Oef. 20)3.35-19 UNITED .STATES NUCLEAR REGULATORY
COMMISSION
WASHINGTON, D. C. 20555 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300 POSTAGE AND FEES PAID WEIED STATES NUCLEAR EGIRA TORY COMMISSION
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Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Plutonium Processing and Fuel Fabrication Plant. (Withdrawn 1/15/98)
ML003739504
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Issue date: 07/31/1979
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RG-3.35, Rev.1
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