Regulatory Guide 3.33
| ML12184A014 | |
| Person / Time | |
|---|---|
| Issue date: | 04/30/1977 |
| From: | Office of Nuclear Regulatory Research |
| To: | |
| References | |
| RG-3.033 | |
| Download: ML12184A014 (21) | |
REGU
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0 U.S. NUCLEAR REGULATORY COMMISSION April 1977
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65.'t REGULATORY GUI LI
OFFICE OF STANDARDS DEVELOPMENT
REGULATORY GUIDE 3.33 1 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL
CONSEQUENCES OF ACCIDENTAL NUCLEAR CRITICALITY IN A FUEL
REPROCESSING PLANT
A. INTRODUCTION
safety research programs. This guide lists assump- tions used to evaluate the magnitude and radiological Section 50.34, "Contents of Applications: consequences of a criticality accident in a fuel Technical Information," of 10 CFR Part 50, "Licens- reprocessing plant.
ing of Production and Utilization Facilities," re- quires that each applicant for a construction permit A criticality accident is an acci esulting in the or operating license provide an analysis and evalua- uncontrolled release of energy o a semblage of tion of the design and performance of structures, fissile material. The circu sta of a ticality ac- systems, and components of the facility with the ob- cident are difficult to wever, the most jective of assessing the risk to public health and safety serious criticality ac nt expected to oc- resulting from operation of the facility and including cur when the reac ty h tent of the deviation determination of the adequacy of structures, systems, from criticalit f ea c ain reacting medium)
and components provided for the prevention of acci- could increas st and without control in dents and the mitigation of the consequences of acci- the fissi c 'on f largest credible mass. In a dents. fuel re p t where conditions that might le c a re carefully avoided because of the In a fuel reprocessing plant, a criticality accident is po* l adverse physical and radiological ef- one of the postulated accidents used to evaluate the c an accident is extremely uncommon.
adequacy of an applicant's proposed activities with e , experience with these and related facilities respect to the public health and safety. The m s monstrated that criticality accidents could oc- described in this guide result from review and tion c on a number of specific cases and, as su fl the latest general NRC-approved approac to the In a fuel reprocessing plant, such an accident might lem. If an applicant desires to e oy in- be initiated by (I) inadvertent transfer or leakage of a formation that may be developed in t or to solution of fissile material from a geometrically safe use an alternative method, NRC will iew the containing vessel into an area or vessel not so proposal and approve ntrke, if found acceptable. designed, (2) introduction of excess fissile material solution to a vessel, (3) introduction of excess fissile material to a solution, (4) overconcentration of a solution, (5) failure to maintain sufficient neutron ab- In the procg of ? w'w applications for permits sorbing materials in a vessel, (6) precipitation of fis- and lice aLhoriz g the construction or opera- sile solids from a solution and their retention in a ves- tion of elg plants, the NRC staff has sel, (7) introduction of neutron moderators or reflec- develoP ppropritely conservative assumptions tors (e.g., by addition of water to a highly under- that are by the staff to evaluate an estimate of moderated system), (8) deformation of or failure to the radiolo cal consequences of various postulated maintain safe storage arrays, or (9) similar actions accidents. These assumptions are based on previous that can lead to increases in the reactivity of fissile accident experience, engineering judgment, and on systems. Some acceptable means for minimizing the the analysis of applicable experimental results from likelihood of such accidents are described in USNRC REGULATORY GUIDES Comments should be sent to the Secretary of the Commission, US. Nuclear Regu- Regulatory Guides are issued to describe and make available to the public methods latory Commission, Washington, D.C. 20555, Attention: Docketing and Service acceptable to the NRC staff of implementing specific parts of the Commission's Branch.
regulations, to delineate techniques used by the staff in evaluating specific problems The guides are issued in the following ten broad divisions or postulated accidents, or to provide guidance to applicants. Regulatory Guides are not substitutes for regulations, and compliance with them is not required. 1. Power Reactors 6. Products Methods and sol utions different from those set Out in the guides will be accept- 2. Research and Test Reactors 7. Transportation able if they provide a basis for the findings requisite to the issuance or continuance 3. Fuels and Materials Facilities 8. Occupational Health of a permit or license by the Commission. 4. Environmental and Siting
9. Antitrust Review
5. Materials and Plant Protection 10. General Comments and suggestions for improvements in these guides are encouraged at all times, and guides will be revised, as appropriate, to accommodate comments and Requests for single copies of issued guides (which may be reproduced) or for place- to reflect new information or experience. However, comments on this guide,if ment on an automatic distribution list for single copies of future guides in specific ieceived within about two months after its issuance, will be particularly useful in divisions should be made in writing to the US. Nuclear Regulatory Commission, evaluating the need for an early revision. Washington, D.C. 20555, Attention: Director. Division of Document Control.
Regulatory Guides 3.4, "Nuclear Criticality Safety in was inversely related to the background neutron flux, Operations with Fissionable Material Outside Reac- which is much greater for plutonium than for tors,"' and 3.1, "Use of Borosilicate Glass Raschig uranium systems.
Rings as a Neutron Absorber in Solutions of Fissile Material."'
1. Criticality Accident Experience in Relation to the Estimation of the Most Severe Accident Those systems consisting only of solid fissile, reflector, or moderator materials exhibited little or no plateau period, whereas solution systems had well developed plateaus. For solution systems, the energy release during the plateau period, because of its dura- ei Stratton (Ref. 1) has reviewed in detail 34 occa- tion, provided the major portion of the total energy sions prior to 1966 when the power level of a fissile released. For purposes of the planning necessary to system increased without control as a result of un- deal adequately with criticality incidents in ex- planned or unexpected changes in its reactivity. perimental and production-type nuclear facilities, Although only six of these incidents occurred in Woodcock (Ref. 3) made use of these data to estimate processing operations, and the remainder occurred possible fission yields from excursions in various mostly in facilities for obtaining criticality data or in types of systems. For example, spike yields of IE+ 17 experimental reactors, the information obtained and and IE+ 18 and total yields of 3E+ 18 and 3E+ 19 fis- its correlation with the characteristics of each system sions were suggested for criticality accidents occur- have been of considerable value for use in estimating ring in solution systems of 100 gallons or less and the consequences of accidental criticality in process more than 100 gallons, respectively. Little or no systems. The incidents occurred in aqueous solutions mechanical damage was predicted at these levels.
of uranium or plutonium (10), in metallic uranium or plutonium in air (9), in inhomogeneous water- 2. Methods Developed for Predicting the Magnitude of moderated systems (9), and in miscellaneous solid Criticality Accidents uranium systems (6).
The nuclear excursion behavior of solutions of The estimated total number of fissions per incident enriched uranium has been studied extensively both ranged from IE+15 2 to IE+20 with a median of theoretically and experimentally. A summary by about 2E+17. More recently another incident in a Dunenfeld and Stitt (Ref. 4) of the kinetic experi- plutonium processing facility in Windscale (U.K.) ments on water boilers, using uranyl sulfate solu- was described in which a total yield of about IE+15 tions, describes the development of a kinetic model fissions apparently occurred (Ref. 2). In ten cases, the that was confirmed by experiment. This model supercriticality was halted by an automatic control device. In the remainder, the shutdown was effected as a consequence of the fission energy release that resulted in thermal expansion, density reduction from defines the effects of thermal expansion and radiolytic gas formation as power-limiting and shut- down mechanisms. 0
the formation of very small bubbles, mixing of light The results of a series of criticality excursion ex- and dense layers, loss of water moderator by boiling, periments resulting from the introduction of uranyl or expulsion of part of the mass. nitrate solutions to vertical cylindrical tanks at vary- ing rates are summarized by Lecorch6 and Seale (Ref.
5). This report confirms the applicability of the kinetics model for solutions, provides correlations of Generally, the criticality incidents were peak power with reactivity addition rate, notes the characterized by an initial burst or spike in the curve importance of a strong neutron source in limiting of fission rate versus time followed by a rapid but in- peak power, and indicates the nature of the plateau complete decay as the shutoff mechanism was in- following the peak.
itiated. As more than one shutdown mechanism may affect the reactivity of the system and the effect of a Many operations with fissile materials in a fuel particular mechanism may be counteracted, the in- reprocessing plant are conducted with aqueous (or itial burst was frequently succeeded by a plateau organic solvent) solutions of fissile materials. Conse- period of varying length. This plateau was quently, well-founded methods for the prediction of characterized by a lesser and declining fission rate total fissions and maximum fission rate for accidents and finally by a further dropoff as shuldown was that might occur in solutions (in process or other ves- completed. The magnitude of the initial burst was sels) by the addition of fissile materials should be of directly related to the rate of increase of reactivity considerable value in evaluating the effects of possi- and its magnitude above the just-critical value but ble reprocessing plant criticality accident
s. From the
' Copies may be obtained from tfie U.S. Nuclear Regulatory results of the excursion studies and from accident Commission, Washington, D.C. 20555, Attention: Director, Divi- data, Tuck (Ref. 6) has developed methods for es- sion of Document Control. timating (1) the maximum number of fissions in a 5-
2 IE+ 15 = I x 10". This notational form will be used throughout second interval (the first spike), (2) the total number this guide.
3.33-2 of fissions, and (3) the maximum specific fission rate I
in vertical cylindrical vessels, 28 to 152 cm in (Ref. 8), and from the SPERT-I reactor transient diameter and separated by > 30 cm from a bottom tests with light- and heavy-water moderated reflecting surface, resulting from the addition of up to uranium-aluminum and U0 2-stainless steel fuels
500 g/l solutions of Pu-239 or U-235 to the vessel at (Ref. 9). Oxide core tests in the latter group provide rates of 0.1 to 7.5 gal/min. Tuck also gives a method some information on energy release mechanisms that for estimating the power level from which the steam- may be effective, for example, in spent fuel storage or generated pressure may be calculated and indicates fuel leaching systems in a reprocessing plant. Review that use of the formulas for tanks >152 cm in of unusual reprocessing structures, systems, and com- diameter is possible with a loss in accuracy. ponents for the possibility of accidental criticality should also consider recognized anomalous situa- Methods for estimating the number of fissions in tions in which the possibility of accidental nuclear the initial burst and the total number of fissions, criticality may be conceived (Ref. 10).
derived from the work reported by Lcorchi and Seale (Ref. 5), have also been developed by Olsen and The application of the double-contingency prin- others (Ref. 7). These were evaluated by application ciple3 to fissile material processing operations has to ten actual accidents which have occurred in solu- been successful in reducing the probability of ac- tions and were shown to give conservative estimates cidental criticality to a low value. As a consequence, in all cases except one. the scenarios required to arrive at accidental criticality involve the assumption of multiple Fission yields for criticality accidents occurring in breakdowns in the nuclear criticality safety controls.
solution and some heterogeneous systems, e.g., li- It has therefore been a practice to simply and conser- quid/fixed geometry, can be reasonably estimated us- vatively assume an accidental criticality of a ing existing methods. However, methods for es- magnitude equal to, or some multiple of, the timating the possible fission yield from other types of historical maximum for all criticality accidents out- heterogeneous systems, e.g., liquid/powder, are less side reactors without using any scenario clearly reliable because of the uncertainties of predicting defined by the specific operations being evaluated. In system reactivity rate. The uncertainties of geometry the absence of sufficient guidance, there has been and moderation result in a broad range of possible wide variation in the credibility of the postulated yields. magnitude of the occurrence (particularly the size of the initial burst), the amount of energy and radioac- Woodcock (Ref. 3) estimated that in solid tivity assumed to be released, and the magnitude of plutonium systems, solid uranium systems, and the calculated consequences.
heterogeneous liquid/powder systems (fissile material not specified) total fission yields (substan- It is the staff's judgment that the evaluation of the tially occurring within the spike) of IE+ 18, 3E+ 19, criticality accident should assume the simultaneous and 3E+20, respectively, could be predicted. breakdown of at least two independent controls Mechanical damage varied from slight to extensive. throughout all elements of the operation. Each con- Heterogeneous systems consisting of metals or solids trol should be such that its circumvention is of very in water were estimated to achieve a possible low probability. Experience has shown that the magnitude of IE+19 following an initial burst of simultaneous failure of two independent controls is
3E+18 fissions. Operations in a fuel reprocessing very unlikely if the controls are derived, applied, and plant involve only a small number of complete as- maintained with a high level of quality assurance.
semblies of fuel rods, except in the fuel storage pool. However, if controls highly dependent on human ac- In the latter area, a rigid array of assemblies is main- tions are involved, this approach will call for some tained and normally only a single assembly may be in variation in the assumed number of control failures.
motion in the vicinity of the array. Consequently, the The criticality accidents so conceived should then be rate of reactivity addition in such a system would be analyzed to determine the most severe within the quite low, and the predicted magnitude of a criticality framework of assumed control failures, using realistic incident would be correspondingly low. These es- values of such variables as the fissile inventory, vessel timates could aid in the analysis of situations in plant sizes, and pump transfer rates.
systems. However, they should not be taken as ab- solute values for criticality assumptions for the pur- 3. Radiological Consequences of Accidestal pose of this guide. Criticality Past practice has been to evaluate the radiological For systems other than solution systems, the es- consequences to individuals of postulated accidental timation of the peak fission rate and the total number criticality in fuel reprocessing plants in terms of a frac- of fissions accompanying an accidental nuclear criticality may be accomplished with the aid of infor- 3 The double-contingency principle is defined in ANSI N16-1-
1969, "Nuclear Criticality Safety in Operations with Fissionable mation derived from accident experience, from ex- Materials Outside Reactors," which is endorsed by Regulatory periments on reactors utilizing bare uranium metal Guide 3.4.
3.33-3
tion of the guideline values in 10 CFR Part 100, ble gases, expected to be present in the spent fuel at
"Reactor Site Criteria." the maximum burnup and the minimum postirradia- tion decay time for which the plant is designed. These The consequences of a criticality accident may be data included in this guide (see Table 1) list the limited by containment, shielding, isolation distance, radioactivity of available significant nuclides assum- or evacuation of adjacent occupied areas subsequent ing 100% dissolution, the burnup to be 33.000
to detection of the accident. If the impact of a criticality MWd/MTU, and a postirradiation decay time of 150
accident is to be limited through evacuation of adja- days.
cent occupied areas, there should be prior, formal ar- rangements with individual occupants and/ The vessel is assumed to be located within a ven- or local authorities sufficient to ensure that such tilated cell which provides shielding equivalent to 5 movements can be effected in the time allowed. feet of concrete with a density of 142 lb/ft3 . The ex-
C. REGULATORY POSITION
cursion produces an initial burst of 1E+ 18 fissions in
0.5 second followed successively at 10-minute inter-
1. Following are the plant assessment and assump- vals by 47 bursts of 1.9E+ 17 fissions for a total of tions related to energy release from a criticality acci- 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 dent and the minimum criticality accident to be con- to be terminated by evaporation of 100 liters of a sidered: solution containing 400 g/l of uranium (<5%
enriched) and concentrations of associated fission a. When defining the characteristics of an assumed products and transuranic elements corresponding to criticality accident in order to assess the adequacy of the sum of those produced in the incident plus those structures, systems, and components provided for the present in irradiated fuel (assuming 100% dissolution)
mitigation of the consequences of accidents, the ap- for the plant design condition. However, the noble plicant should evaluate credible criticality accidents gas fission products initially present in the fuel are as- in all those elements of the plant provided for the sumed to have been removed prior to the incident.
storage, handling, or processing of fissile materials or Table 2 lists the radioactivity of significant nuclides into which fissile materials in significant amounts released from the criticality accident.
could be introduced. To determine the circumstances of the criticality accidents, controls judged equivalent 2. Assumptions related to the release of radioactive to at least two highly reliable, independent criticality material are as follows:.
controls should be assumed to be circumvented. The a. It should be assumed that all of the noble gas 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 fission products (except those removed prior to the excursion), 25% of the iodine radionuclides, and 0.1%
of the ruthenium radionuclides resulting from the ex- a
accident should then be selected for the assessment of cursion or initially present in the spent fuel are the adequacy of the plant. released directly to the cell atmosphere. It should also be assumed that an aerosol, which is generated from Calculation of the radioactivity of fission products the evaporation of solution during the excursion, is and transuranic elements initially present and later released directly to the cell atmosphere. The aerosol produced in the incident should be accomplished by should be assumed to comprise 0.05% of the salt con- computer codes ORIGEN (Ref. 11) and RIBD (Ref. tent of the solution that is evaporate
d. The cell
12), respectively. An equivalent calculation may be volume and ventilation rate should be considered on substituted, if justified on an individual case basis. an individual case basis.
b. If the results of the preceding evaluation in- b. The effects of radiological decay during transit dicate that no possible criticality accident exceeds in in cell and in the plant exhaust system should be severity the criticality accident postulated in this sec- taken into account on an individual case basis.
tion, then the conditions of the following example c. The reduction in the amount of radioactive may be assumed for the purpose of assessing the ade- quacy of the facility. A less conservative set of condi- material available for release to the environment tions may be used if they are shown to be applicable through the plant stack(s) as a result of the normal by the specific analyses conducted in accordance with operation of sorption or filtration systems in the paragraph C.l.a above. plant exhaust systems may be taken into account, but the amount of reduction in the concentration of An excursion is assumed to occur in a vented vessel radioactive materials should be evaluated on an in- of unfavorable geometry containing a solution of 400 dividual case basis.
g/l of uranium enriched to less than 5% U-235. The ' Certain assumptions for release of radioactive material, dose solution is also assumed to contain all of the trans- conversions, and atmospheric diffusion reflect the staff's position uranic elements and fission products, except the no-
3.33-4 indicated in Regulatory Guide 1.3 (Ref. 22).
I
3. Acceptable assumptions for dose and dose conver- m3/sec. These values were developed from the sion are as follows: average daily breathing rate (2E + 7 cm 3 /day) as- sumed in the report of ICRP Committee 11-1959 a. The applicant should show that the conse- (Ref. 14).
quences of the prompt gamma and neutron dose are sufficiently mitigated to allow occupancy of areas d. External whole body doses should be calculated necessary to maintain the plant in a safe condition using "infinite cloud" assumptions, i.e., the dimen- following the accident. The following semi-empirical sions of the cloud are assumed to be large compared equations should be used for these calculations. to the distance that the gamma rays and beta particles These equations are acceptable to the NRC staff and travel. "Such a cloud would be considered an infinite were developed from experimental data. Different cloud for a receptor at the center because any ad- methods may be substituted, if justified on an in- ditional [gamma and] beta emitting material beyond dividual case basis. Potential total dose attenuation the cloud dimensions would not alter the flux of due to shielding and dose exposures should be [gamma rays and] beta particles to the receptor."
evaluated on an individual case basis. [See Meteorology and A tomic Energy-1968 (Ref. 15),
(1) Prompt' Gamma Dose Section 7.4.1.1; editorial additions made so that gam- ma and beta emitting material could be considered.]
Dy = 2.1E-20N d2 e-3.4d Under these conditions the rate of energy absorption where per unit volume is equal to the rate of energy released per unit volume. For an infinite uniform cloud con- Dy = gamma dose (rem) taining x curies of beta radioactivity per cubic meter, N = number of fissions the beta dose rate in air at the cloud center is d = distance from source (km)
PDz0 0.457 Ep X
Data presented in The Effects of Nuclear Weapons (Ref. 13, p. 384) should be used to develop The surface body dose rate from beta emitters in the dose reduction factors. For concrete, the dose should infinite cloud can be approximated as being one-half be reduced by a factor of 2.5 for the first 8 inches, a this amount (i.e., 3D-* = 0.23 Ep3x). For gamma factor of 5.0 for the first foot, and a factor of 5.5 for emitting material, the dose rate in air at the cloud each additional foot. center is YD = 0.507 EyX
(2) Prompt Neutron Dose
2 From a semi-infinite cloud, the gamma dose rate in Dn = 7E- 20N d" e-1.2d air is where yD * = 0.25 EX
Dn = neutron dose (rem) where N = number of fissions PD:ý = beta dose rate from an infinite cloud d = distance from source (km) (rad/sec)
For concrete, the dose should be reduced by a yDc- = gamma dose rate from an infinite cloud factor of 2.3 for the first 8 inches, 4.6 for the first (rad/sec)
foot, and a factor of 20 for each additional foot. E13 = average beta energy per disintegration (MeV/dis)
b. No correction should be made for depletion of = average gamma energy per disintegration the effluent plume of radioactive iodine due to (MeV/dis)
deposition on the ground or for the radiological decay of iodine in transit. X = concentration of beta or gamma emitting isotope in the cloud (Ci/m 3 )
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 e. The following specific assumptions are accep- person offsite should be assumed to be 3.47E-4 table with respect to the radioactive cloud dose m3/sec. Frorm 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, calculations:
the breathing rate should be assumed to be 1.75E-4 I Most of the neutron and part of the gamma radiation are emitted (1) The dose at any distance from the plant in the actual fission process. Some gamma radiation is produced in should be calculated based on the maximum con- various secondary nuclear processes, including decay of fission products. For the purposes of this guide, "prompt" gamma doses centration time integral (in the course of the accident)
should be evaluated including the effects of decay of significant fis- in the plume at that distance, taking into account sion products during the first minute of the excursion. specific meteorological, topographical, and other
3.33-5
characteristics that may affect the maximum plume a. Elevated releases should be considered to be at a6 concentration. These site-related characteristics height equal to no more than the actual stack height.
should be evaluated on. an individual case basis. In Certain site-dependent conditions may exist, such as the case of beta radiation, the receptor is assumed to surrounding elevated topography or nearby struc- be exposed to an infinite cloud at the maximum tures, that will have the effect of reducing the actual ground level concentration at that distance from the stack height. The degree of stack height reduction plant. In the case of gamma radiation, the receptor is should be evaluated on an individual case basis.
assumed to be exposed to only one-half the cloud ow- ing to the presence of the ground. The maximum Also, special meteorological and geographical con- cloud concentration should always be assumed to be ditions may exist which can contribute to greater at ground level. ground level concentrations in the immediate neighborhood of a stack. For example, fumigation
(2) The appropriate average beta and gamma should always be assumed to occur; however, the energies emitted per disintegration used should be as length of time that a fumigation condition exists is given in the Table of Isotopes (Ref. 16). strongly dependent on geographical and seasonal fac- tors and should be evaluated on an individual case basis.7 (See Fig. 3 for elevated releases under fumiga-
(3) The whole body dose should be considered tion conditions.)
as the dose from gamma radiation at a depth of 5 cm and the genetic dose at a depth of I cm. The skin dose b. For plants with stacks, the atmospheric diffu- should be the sum of the surface gamma dose and the sion model should be as follows:
beta dose at a depth of 7 gm/cm2 . The beta skin dose may be estimated by applying an energy dependent attenuation factor (Dd/DB) to the surface dose (1) The basic equation for atmospheric diffusion according to a method developed by Loevinger, from an elevated release is Japha, and Brownell (Ref. 17). (See Figure 1.)
22 2 f. The "critical organ" dose from the inhaled X/Q =r~exp(-he / oz )
radioactive materials should be estimated. The
"critical organ" is that organ which receives the highest radiation dose after the isotope is absorbed where into the body. For the purpose of this guide, the fol- lowing assumptions should be made:
X = the short-term average centerline value of the ground level concentration (Ci/m 3)
I
(1) The radionuclide dose conversion factors are Q = amount of material release (Ci/sec)
as recommended by the report of Committee I1, ICRP (Ref. 14). tL = windspeed (m/sec)
U = the horizontal standard deviation of the plume Y (meters). (See Ref. 21, Figure V-1, p.48.)
(2) The effective half-life for the nuclide is as recommended in ICRP Publication 6 (Ref. 18). z= the vertical standard deviation of the plume (meters). (See Ref. 21, Figure V-2, p.48.)
he = effective height of release (m) 8
(3.) The plutonium and other actinide nuclide
6 Credit for an elevated release should be given only if the point of clearance half time, or fraction of nuclide clearing the release is (I) more than two and one-half times the height of any organ, is as recommended by the ICRP task group on structures close enough to affect the dispersion of the plume or (2)
lung dynamics (Ref. 19). A computer code, located far enough from any structure that could have an effect DACRIN, (Ref. 20) is available for this model. Task on the dispersion of the plume. For these plants without stacks, the group lung model (TGLM) clearance parameters are atmospheric diffusion factors assuming ground level releases, as presented in Table 3; the model is shown schematical- shown in Regulatory Position 4.c, should be used.
ly in Figure 2. ' 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 g. The potential dose for all significant nuclides for one-half hour. For sites located less than 2 miles from large should be estimated for the population distribution bodies of water, a fumigation condition should be assumed to exist on a site-related basis. at the time of the accident and continue for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
I he = h, - ht, where h, is the height of the release above plant grade and ht is the maximum terrain height, above plant grade,
4. Acceptable assumptions for atmospheric diffusion between the point of release and the point at which the calculation are as follows:
4 is made. ht should not be allowed to exceed h,.
3.33-6
(2) For time periods of greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the where plume from an elevated release should be assumed to X = the short-term average centerline value of the meander and spread uniformly over a 22.50 sector.9 3 ground level concentration (Ci/.m )
The resultant equation is Q = amount of material release (Ci/sec)
2.032 exp( -h2 /2oz.2) ýL = windspeed (m/sec)
x/Q = aztx Cr = the horizontal standard deviation of the plume where (m). (See Ref. 21, Figure V-I, p.48.)
x = distance from the release point (meters); other 017 = the vertical standard deviation of the plume variables are as given in b(l). (m). (See Ref.21, Figure V-2, p.48.)
(3) The atmospheric diffusion model'" for an (3) For time periods of greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the elevated release as a function of the distance from the plume should be assumed to meander and spread un- plant is based on the information in the table below iformly over a 22.5' sector.' The resultant equation is Time Following 2.032 x/Q- =zp-ZX
Accident Atmospheric Conditions
0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> See Figure 4 for Envelope of Pasquill diffusion categories where
[based on Figure A7, Meteorology and Atomic x = distance from point of release to the receptor;
Energv-1968 (Ref. 15), as- other variables are as given in c(2).
suming various stack heights]
windspeed I m/sec; uniform (4) The atmospheric diffusion model for ground direction. level releases is based on the information in the fol- lowing table:
8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> See Figure 5 for Envelope of Pasquill diffusion categories; Time Following windspeed 1 m/sec; variable Accident Atmospheric Conditions direction within a 22.50 sector.
0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Pasquill Type F, windspeed I
c. For facilities exhausted without stacks, the at- m/sec, uniform direction mospheric diffusion model should be as follows: 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 I
m/sec, variable direction
(1) The 0-to-8 hour ground level release con- within a 22.50 sector.
centrations may be reduced by a factor ranging from one to a.maximum of three (see Figure 6) for ad- (5) Figures 7A and 7B give the ground level ditional dispersion when calculating nearby potential release atmospheric diffusion factors based on the exposures. The volumetric building wake correction parameters given in c(4).
factor, as defined in Section 3-3.5.2 of Meteorology and A tomnic Energy-1968 (Ref. 15), should be used in the 0-to-8 hour period only; it is used with a shape
D. IMPLEMENTATION
factor of one-half and the minimum cross-sectional area of a major building. .The purpose of this section is to provide informa- tion to applicants and licensees regarding the staffs
(2) The basic equation for atmospheric diffusion plans for using this regulatory guide.
from a ground level point source is Except in those cases in which the applicant proposes an alternative method for complying with I specified portions of the Commission's regulations, x/Q = --
___
the method described herein will be used in the evaluation of submittals for operating license or con-
9 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 struction permit applications docketed after meteorological data are available to justify a wind direction December 1, 1977.
change. This should be considered on an individual case basis.
'0 In some cases, site-dependent parameters such as meteorology, If an applicant wishes to use this regulatory guide topography, and local geography may dictate the use of a more in developing submittals for applications docketed on restrictive model to ensure a conservative estimate of potential off- site exposures. Site-related meteorology should be developed on an or before December 1, 1977, the pertinent portions individual case basis. If adequate local meteorological data are not of the application will be evaluated on the basis of available, this model should be used. this guide.
3.33-7
REFERENCES
1. W. R. Stratton, "Review of Criticality Incidents," 13. The Effects of Nuclear Weapons, Revised Ed.,
LA-361 1, Los Alamos Scientific Laboratory (Jan. Samuel Glasstone, Editor, U.S. Depar
t. of Defense
1967). (Feb. 1964).
2. T. G. Hughes, "Criticality Incident at Windscale," 14. "Permissible Dose for International Radiation,"
Nuclear Engineering International,Vol. 17, No. 191, Publication 2, Report of Committee II, International pp.95-7 (Feb. 1972). Commission on Radiological Protection (ICRP),
3. E. R. Woodcock, "Potential Magnitude of Pergamon Press (1959).
Criticality Accidents," AHSB(RP)R-14, United 15. Meteorology and Atomic Energy-I1968, D. H.
Kingdom Atomic Energy Authority. Slade, Editor, U.S. Atomic Energy Commission (July
4. M.S. Dunenfeld, R. K. Stitt, "Summary Review 1968).
of the Kinetics Experiments on Water Boilers,"
16. C. M. Lederer, J. M. Hollander, I. Perlman, NAA-SR-7087, Atomic International (Feb. 1973).
Table of Isotopes, 6th Ed., Lawrence Radiation
5. P. Lcorch6, R. L. Seale, "A Review of the Experi- Laboratory, Univ. of California, Berkeley, CA.
ments Performed to Determine the Radiological Consequences of a Criticality Accident," Y-CDC-12, 17. Radiation Dosimetry, G. J. Hine and G. L.
Union Carbide Corp. (Nov. 1973). Brownell, Editors, Academic Press, New York
(1956).
6. G. Tuck, "Simplified Methods of Estimating the Results of Accidental Solution Excursions," Nucl. 18. Recommendations of ICRP, Publication 6, Technol., Vol. 23, p. 17 7 (1974). Pergamon Press (1962).
7. A. R. Olsen, R. L. Hooper, V. 0. Uotinen, C. L. 19. "The Metabolism of Compounds of Plutonium Brown, "Empirical Model to Estimate Energy and Other Actinides," a report prepared by a Task Releases from Accidental Criticality," ANS Trans., G -- of Committee 1I, ICRP, Pergamon Press Vol. 19, pp. 189-91 (1974). tMay 1972).
8. T. F. Wimmette et al., "Godiva 2-An Un- moderated Pulse Irradiation Reactor," Nucl. Sci. 20. J. R. Houston, D. L. Strenge, and E. C. Watson, Eng., Vol. 8, p. 6 9 1 (1960). "DACRIN-A Computer Program for Calculating Organ Dose from Acute or Chronic Radionuclide
9. W. E. Nyer, G. 0. Bright, R. J. McWhorter,
"Reactor Excursion Behavior," International Conference on the Peaceful Uses of Atomic Energy, Inhalation," BNWL-B-389(UC-4), Battelle Memorial Institute, Pacific Northwest Laboratories, to Richland, WA.(Dec.1974).
paper 283, Geneva (1966).
21. F. A. Gifford, Jr., "Use of Routine
10. E. D. Clayton, "Anomalies of Criticality," Nucl. Meteorological Observations for Estimating At- Technol., Vol. 23, No. 14 (1974). mospheric Dispersion," Nuclear Safety, Vol. 2, No.
11. M. J. Bell, "ORIGEN-The ORNL Isotope 4, p.48 (June 1961).
Generation and Depletion Code," ORNL-4628, Oak
22. Regulatory Guide 1.3, "Assumptions Used for Ridge National Laboratory (May 1973). Evaluating the Radiological Consequences of a Loss
12. R. 0. Gumprecht, "Mathematical Basis of Com- of Coolant Accident for Boiling Water Reactors,"
puter Code RIBD," DUN-4136, Douglas United U.S. Nuclear Regulatory Commission, Washington, Nuclear, Inc. (June 1968). D.C.
3.33-8 l
TABLE 1 ASSUMED FISSION PRODUCT AND TRANSURANIC
NUCLIDE RADIOACTIVITY IN SPENT FUEL SOLUTION
PRIOR TO CRITICALITY INCIDENT
3.3% Enriched Fuel Irradiated to 33000 MWd/MTU,
cooled 150 days and calculated by ORIGEN code.
NUCLIDE CURIES/LITER
Tritium 2.9E- I
Strontium-89 4.0E+ 1 Strontium-90 3.2E+1 Yttrium-90 3.2E+ I
Yttrium-991 *5.7E+ I
Zirconium-95 1.2E+2 Niobium-95 2.2E+2 Ruthenium- 103 3.7E+ 1 Rhodium-103M 3.7E+ I
Ruthenium-106 1.7E+2 Rhodium-106 1.7E+2 Iodine-129 1.6E - 5 Iodine-131 9.1E -4 Xenon-131m .4E - 3 Cesium- 139 9.OE+ I
Cesium- 137 4.5E+1 Barium- 137M 4.2E+ I
Cerium-141 2.4E+ I
Cerium-144 3.2E+2 Praseodymium-144 3.2E+2 Promethium- 147 4.2E+ I
Europium- 154 2.3E 0
Plutonium-238 1.2E 0
Plutonium-239 1.4E- I
Plutonium-240 2.OE - 1 Plutonium-241 4.8E+I
Americium-241 8.4E - 2 Curium-242 6.3E 0
Curium-244 L.OE 0
3.33-9
TABLE 2 RADIOACTIVITY OF IMPORTANT NUCLIDES RELEASED
FROM THE CRITICALITY ACCIDENT IN THIS GUIDE (Ci)
NUCLIDE 0 to 0.5 hr 0.5 to 8 hr
3.3E+ 1 TOTAL
3.7E+ I
0
Kr-83m 3.7E 0
Kr-85m 1.6E+ 1 1.5E+2 1.7E+2 Kr-85 1.5E-4 1.4E-3 1.6E-3 Kr-87 1.0E+2 9.OE+ 2 1.0E+3 Kr-88 6.5E+ 1 5.9E+2 6.6E+2 Kr-89 4.1E+3 3.7E+4 4.1E+4 Xe-131m 3.8E-4 3.5E-3 3.9E-3 Xe-133m 5.5E-2 4.9E-1 5.5E-1 Xe-133 1.3E 0 1.2E+ I 1.3E+ 1 Xe-135m 1. IE+ 1 9.9E+1 1.1E+2 Xe-135 1.6E+ 1 1.5E+2 1.7E+2 Xe-137 3.8E+3 3.5E+4 3.9E+4 Xe-138 1.2E+3 1.0E+4 1.IE+4
1-129 4.2E-I1 3.9E-10 4.3E-10
1-131 1.8E-1 1.6E0 1.8E0
1-132 6.7E-1 6.1E0 6.7E0
1-133 3.5E0 3.1E+I 3.5E+1
1-134 4.8E+ I el '-+2 4.8E+2
1-135 1.2E+1 1.OE+2 1.2E+2
3.33-10
TABLE 3 VALUES OF THE CLEARANCE PARAMETERS FOR THE
TASK GROUP LUNG MODELa COMPA RTMENT CLASS Db, c CLASS Wc CLASS yc NP Tk d fkd Tkd Tkd fk d fkd a 0.01 0.5 0.01 0.1 0.01 0.01 b 0.01 0.5 0.4 0.9 0.4 0.99 TB c 0.01 0.95 0.01 0.5 0.01 0.01 d 0.2 0.05 0.2 0.5 0.2 0.99 P e 0.5 0.8 50 0.15 500 0.05 f n.a.e n.a.e 1.0 0.4 1.0 0.4 g n.a.e n.a. e 50 0.4 500 0.4 h 0.5 0.2 50 0.05 500 0.15 L i 0.5 1.0 50 1.0 1000 0.9 a See Figure 2 for the task group lung model (TGLM) schematic diagram.
b Data for soluble plutonium are included. To maintain dose conversion conservatism, this class should only be considered if justified on an individual case basis.
c Class D = readily soluble compounds where removal time is measured in days.
Class W = compounds with limited solubility where removal time is measured in weeks.
d Class Y = insoluble compounds where removal time is measured in years.
T
k 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 I 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.
e n.a. means not applicable.
3.33-11
1.0
,00 :41 1 / 01 f-~-
'": IY /
10 -1
0i /1 ii /0i54
0c
102 il
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 Plane Source of Infinite Thickness and for Allowed Spectra Developed from Considerations Presented in Reference 17, Chapter 16 FIGURE 1
3.33-12 I
I j LYMPH I
SCHEMATIC DIAGRAM DEVELOPED FROM ICRP TASK GROUP LUNG MODEL (Ref. 19)
FIGURE 2
3.33-13
as . Ow I .
110-3 1 i------
- -T
.... i.,
lip.I ..
.2
' *h~lO0 meters.
S10-
10-4 10 10
10 102 10
A.i
104 j ; j:A iliý
105 Distance from Release Point (meters)
FIGURE 3(Ref. 22)
U
3.33-14
.... * . .i- :-'" i_ .. . .. 7 "- . ... 7
"~ l 7--[
....... --
_.._* !* -72¶:7.** _ T7-7-.K.
-.
I -7, --
ELEVATED ýRELEASE i[..
- --*"*-7 ........
I ATMOSPHERIC DIFFUSION FACTORS
l0-8 HOUR RELEASE TIME
l 5O 11 1 1 i r -ý 1 1ý% --.* -II ,
4r --
Sh=125 meters*,- 1. 1 L.
Uo--
C -Ihl5metersi.
- -
-I- 7-A
10-8
102 103 104 Distance from Release Point (meters)
FIGURE 4(Ref. 22)
3.33-15
l0
C4 a. Lill 'P,,M f ý1WM.
ILLI
h=t50 m*trs
10-
M.- Tt N.... ..
0-74 =25m DistaefrmRlaePit(ters~
10 - 1 41fIGURET 5(r. 22)L
114 li p.
102 0 104 i jI iHU 105 Distance ~glu fro ReesePinimees F~IGH. R I ef 2 t~t~th
3.33-16
V1 U S!
I
I.
I I I I I I
tItil i J* 01 I
- .1- !l:ll!l:l
~.i. I.
- I
I I. ... ,:,,:I;vI::;.;: :1;
I: ii, 4 V
Til
\
3 1- !1.11 IT Ma.
Ii
,, !i H+/-F
F1
________________________________.11__ 11!! 111 P1I
\I \\\\i I]
11fl Ii BUILDING WAKE
CORRECTION FACTOR' i T
2.5
,!,,,.
- 0.5A=500 meters2/ *'". *
- -- 1--*-_5-1 meters 0.5A=250-0mete'rs"V-
2 S0.5A=1500- me -OloA=3o00 meter7 L1 K 2
0
0 .5A=2000 meters IiI I-h-I ir~ -h C
1.5
- 01 NN
r f+I -viDO.
0.5 vizi~ i -7LL*
Lvi L!
7- J
I
I. *'-'i '*!*
-HL T:7 T"i
.1 Mil-Fr +4
2
10 103 Distance from Structure (meters)
FIGURE 6(Ref. 22)
F
10- 2 k:
I L
, -- ..
L.
I .
I I I II III I111 IIEThEfEFEF3IIW
I ; -II
I ,- I I I i.
1 11 1 1 1 .
1 1 .1 1 1 1 1 f [
I I 1LEVEL
1 llIIRELEASE
i I
1 1GROUND 9LFL
b * [ I *
1144144J
[i!:*,i!i: ATMOSPHERIC DIFFUSION FACTORS FOR -- it I it ,
N VARIOUS TIMES FOLLOWING ACCIDENT
- -:- ] " *. " ]*.1 N
-4---F* 'Fin N=.
-h
....- tT
F-.-
R
10- 3
-T-H
- L.
it - i:P7
-L
H 8--24 hours L1112 U.
U-
L_ -- It ._1..
- -- - 1~~*~~
.2 I
T II
N I I I
-j
10-5 I LW
10 2 104 Distance from Structure (meters)
FIGURE 7A(Ref. 22)
3.33-18 U
10-5
-1--x 1127 I I *I I I I
th I f .
GROUND LEVEL RELEASE
-ATMOSPHERIC DIFFUSION FACTORS FOR
- VARIOUS TIMES FOLLOWING ACCIDENT
N .I III
I 1-T_
N 1-8 hours
-24hot rs S
10-6 N
-t K4' IEH7
_7
-1--I
-F
IIIL
-~
2 I--
U.
- zt--n , 3' i
!----
HF
7-
,-i -- i- i -+- +-
'2- I
1
1 1 1 1 1 11 Il I I I . -1-1-1 1ITT 1 1 1ý I 1 1 1 11 L-7. 1
__
I
.1
... .
. . --
I---
.
ii 1%
-Ijjl---iF -Ii I -
10-8 I I* II I .1.1.1 .....~........F..i
10E3 Distance from Structure 106 (meters)
FIGURE 7B(Ref. 22)
3.33-19
UNITED STATES
NUCLEAR REGULAfORY COMMISSION
WASHINGTON. D. C. 20565 POSTAGE AND
U.S. NUCLEAR FEES PAID
REGULATORY
OFFICIAL BUSINESS COMMISSION
PENALTY FOR PRIVATE USE, $300
19406002001 U S NRC SF
0477 OFFICE OF INSPECTION
R J BORES & ENFORCE
631 PARK AVENUE
KING OF PRUSSIA
PA 19406 A
1/8/73 U.S. ATOMIC ENERGY COMMISSION
REGULATORY
DIRECTORATE OF REGULATORY STANDARDS
GUIDE
REGULATORY GUIDE 3.3 QUALITY ASSURANCE PROGRAM REQUIREMENTS
FOR FUEL REPROCESSING PLANTS
A. INTRODUCTION
This standard is adaptable to fuel reprocessing plants as it was prepared to satisfy the. intent and amplify the Appendix B to 10 CFR Part 50, "Quality Assurance requirements of AEC quality assurance regulations and Criteria for Nuclear Power Plants and Fuel Reprocessing provides general requirements and guidance for Plants." establishes quality assurance requirements for establishment and execution of quality assurance the design., construction, and operation of nuclear power programs.
plant and fuel reprocessing plant structures, systems and comtxmnents. This regulatory guide describes an acceptable method of complying with the Commission's
C. REGULATORY POSITION
regulations with regard to overall quality assurance program requirements for fuel reprocessing plants.
The general requirements and guidelines for
B. DISCUSSION
establishing and executing quality assurance programs for nuclear power plants which are included in ANSI
N45.2-197 I, "Quality Assurance Program Requirements Subcommittee N45-3, Nuclear Quality Assurance Standards, (formerly ad hoc Committee N45-3.7) of the for Nuclear Power Plants" may be adapted to fuel American National Standards Institute, Standards reprocessing plants and are generally acceptable and Committee N45. Reactor Plants and Their Main tenance, provide an adequate basis for complying with the program requirements of Appendix B to 10 CFR Part under the sponsorship of the American Society of Mechanical Engineers, has developed a standard which 50, applicable to fuel reprocessing plants. When adapting includes general requirements and guidance for the ANSI N45.2-1971 to fuel reprocessing plants the term establishment and execution of quality assurance fuel reprocessing plants should be substituted as programis during the design, construction, and operation applicable wherever specific reference is made to the phases of nuclear power plants. This standard was term nuclear power plant(s).
approved by the American National Standards Committee N45 and its Secretariat, and it was subsequently approved and designated N45.2-1971 by
'Copies may be obtained from the American Society of the American National Standards Institute on October Mechanical Engineers. United Engineering Center, 345 East 47th
20. 1972. Street, New York, N.Y. 10017.
USAEC REGULATORY GUIDES Copies of published guides may be obtained by request indicating the divisions desired to the US. Atomic Energy Commission, Washington, D.C. 20545, Regulatory Guides are issued to describe and make available to the public Attention: Director of Regulatory Standards. Comments and suggestions for methods acceptable to the AEC Regulatory staff of implementing specific parts of improvements in these guides are encouraged and should be sent to the Secretary the Commission's regulations, to delineate techniques used by the staff in of the Commission, U.S. Atomic Energy Commission, Washington, nDC. 20545, evaluating specific problems or postulated accidents, or to provide guidance to Attention: Chief, Public Proceedings Staff.
applicants. Regulatory Guides are not substitutes for regulations and compliance with them is not required. Methods and solutions different from those set Out in The guides are issued in the following ten broad divisions:
the guides will be acceptable if they provide a basis for the findings requisite to the issuance or continuance of a permit or license by the Commission. 1. Power Reactors
6. Products
2. Research and Test Reactors
7. Transportation
3. Fuels and Materials Facilities S. Occupational Health Published guide5 will be revised periodically, as appropriate, to accommodate 4. Environmental and Siting 9. Antitrust Review comrrsetts and to reflect new information or experience. 6. Materials and Plant Protection 10. General