Regulatory Guide 1.4

Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors

ML13350A195
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Issue date:	06/30/1973
From:	US Atomic Energy Commission (AEC)
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RG-1.004, Rev. 1
Download: ML13350A195 (6)

Revision 1 Revision 1 June 1973 U.S. ATOMIC ENERGY COMMISSION

REGULATORY

DIRECTORATE OF REGULATORY STANDARDS

GUIDE

REGULATORY GUIDE 1.4 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES

OF A LOSS OF COOLANT ACf',DENT FOR PRESSURIZED WATER REACTORS'

C. REGULATORY POSITION

Sect ion 50.34 o1f 10 CFR Pairl 50 requires that each

1. The assuimptions related io the release of radioactive applicant fir a c(nstruiction permit or operating license material from the fuel and containment are as Ibllows:

provid,: an analysis and cvalua3ion of the design and a. T we n t y -five percent of the equilibriut ierlo*rmance of structures. systems, and components of radioactive iodine inventory developed from imlaximu i tile facility with [he objective of assessing fhe risk to full power operation of the core should be assumtned to public health and safety resulting froim operation of the be immediately available for leakage from the prinmary facility. Tile design basis loss of" coolant accident reactor containment. Ninety-one percent of this 25 (LOCA) is one of the postulated accidents Used to percent is to be assumed ito he ill Ithe forma ofelenllelllal evaluate the adequacy of these structures, systems. and iodine. 5 percent of this 25 percent ill the form of comiponents with respect to the public ltealth and safety. particulate iodine. and 4 percent of this 25 percent in This guide gives acceptable assumptions that may be the form of organic iodides.

used in evaluating tIle radiologcal consequences of this b. One hundred percent of the equilibrium accident for a pressurized water reactor. In some cases. radioactive noble gas inventory developed front unusual site characteristics, platit design features. or maximum full power operation od the core should be other factors may require different assumptions which assumed to be immediately available for leakage front will be considered on an individual case basis. The the reactor containment.

Advisory Committee on Reactor Safeguards has been c. The effects of radiological decay during holdup consulted concerning this guide and has concurred in the in the containment or other buildings should be taken regulatory position. into account.

B. DISCUSSION

d. The reduction in the amotunt of radioactive material available for leakage to tile environment by After reviewing a number of applications for containment sprays, recirculating filter systems, or other construction permits and operating licenses for engineered safety features may be taken into account.

pressurized wateli power reactors, the AEC Regulatory but the amount of reduction in concentration of staff has developed a number of appropriately radioactive materials should be evaluated on an conservative assumptions, based on engineering individual case basis.

judgment and on applicable experimental results from e. The primary reactor containment should be safety research programs conducted by the AEC and the assumed to leak at the leak rate incorporated or to le nuclear industry, that are used to evaluate calculations incorporated as a technical specification requirement at of the radioloocal consequences of various postulated peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. and at 50

acciden ts. percent of this leak rate for the remaining duration of the accideint. 2 Peak accident pressure is the maximum1 This guide lists acceptable assumptions that may be pressure defined in the technical specifications for used to evaluate the design basis LOCA of a Pressurized containment leak testing.

Water Reactor (PWR). It should be shown that thc

2 offsite dose consequences will be within thie guidelines Thte effect on coniainnmeni leakage tinder accident of 10 CFR Part 100, conditions of features provided to reduce the leakage ot"

radioactive materials from the containment will be evaluated on

'This guide is a revision of former Safety Guide 4. an individual case basis.

USAEC REGULATORY GUIDES Coples of published guldes may be obtained by request Indicating the divisions desired to the US. Atomic Energy Commission. Washington. 0.1, 20545, Regulatory Guides are issued to describe and make avaliable to the public Attention: Director of Regulatory Standards. Comments and tuggrsilons for methods acceptable to the AEC Regulatory staff of Implementing specific parts of impfrovements In these guides ere encouraged end should be sent to the Secretary the Commission's regulations, to delineate techniques used by the staff in of the Commission, US. Atomic Energy Commission, Washington. O.C. 20545.

evaluating specific problems or postulated accid3nts. 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 solutlons different from those set out in The guides are issued In the following ten broad divliions:

the guides will be acceptable if they provide a basis for the findings requisite to 8. Products the issuance or continuance of a permit or license by the Comrrssio

1. Power Reactors

2. Researcha nd Tast Reactors

7. Transportation

3. Fuels and Materials Facilities 8. Occupational Health Published guides will be revised periodically, as appropriate, to accommodate 4. Environmental end Siting 9. Antlitrust Review comments and to reflect new informatio" or experience. 5. Materials and Plant Protection 1

0. General

.1

2. Acceptable assumptions for atmospheric diffusion From a semi-infinite cloud, the gamma dose rate in air and dose conversion are: is:

a. The 0-8 hour ground level release concentrations may be reduced by a factor ranging from ,D = 0,25E

one to a maximum of three (.see Figure I) for additional dispersion produced by the turbulent wake of the Where reactor building in calculating potential exposures. The volumetric building wake correction, as defined in beta dose rate from an infinite cloud (rad/sec)

section 3.3.5.2 of Meteorology and Atomic Energy gamma dose rate from an infinite cloud

1968. should be used only in the 0-8 hour period: it is (rad/sec)

used with a shape factor of 112 and the minimum EO3= average beta energy per disintegration cross-sectional area of the reactor building only. (Mev/dis)

b. No correction should be made for depletion of' E = average gamma energy per disintegration the effluent plume of radioactive iodine due to (Mev/dis)

deposition on the ground, or for the radiological decay X = concentration of beta or gamma emilling of iodine in transit. isotope in the cloud (curie/m3)

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 persons offsite should be assumed to be 3.47 x 10' f. The following specific assumptions are cubic meters per second. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following acceptable with respect to the radioactive cloud dose the accident, the breathing rate should be assumed to be calculations:

1.75 x 104 cubic meters per second. After that until the (1) The dose at any distance from the reactor end of the accident, the rate should be assumed to be should be calculated based on the maximum

2.32 x 104 cubic meters per second. (These values were concentration in the plume at that distance taking into developed from the average daily breathing rate [2 x 107 account specific meteorological, topographical, and cnv'/dayJ assumed in the report of ICRP, Committee other characteristics which may affect the maximum

11-1959.) plume concentration. These site related characteristics d. The iodine dose conversion factors are given in must be evaluated on an individual case basis. In the case ICRP Publication 2, Report of Committee 11, of beta radiation, the receptor is assumed to be exposed

"Permissible Dose for Internal Radiation," 1959. to an infinite cloud at the maximum ground level e. External whole body doses should be calculated concentration at that distance from the reactor. In the using "Infinite Cloud" assumptions, i.e., the dimensions case of gamma radiation, the receptor is assumed to be of the cloud are assumed to be large compared to the exposed to only one-half the cloud owing to the distance that the gamma rays and beta particles travel. presence of the groun

d. The maximum cloud

"Such a cloud would be considered an infinite cloud for concentration always should be assumed to be at ground a receptor at the center because any additional [gamma level.

and] beta emitting material beyond the cloud (2) The appropriate average beta and gamma dimensions would not alter the flux of [gamma rays energies emitted per disintegration, as given in the Table and] beta particles to the receptor" (Meteorology and of Isotopes, Sixth Edition, by C. M. Lederer, J. M.

Atomic Energy, Section 7.4. .1.-editorial additions Hollander, I. Perlman; University of California, Berkeley, made so that gamma and beta emitting material could be Lawrence Radiation Laboratory; should be used.

considered). Under these conditions the rate of energy absorption per unit volume is equal to the rate of energy g. The atmospheric diffusion model should be as released per unit volume. For an infinite uniform cloud follows:

containing X curies of beta radioactivity per cubic meter (1) The basic equation for atmospheric the beta dose in air at the cloud center is: diffusion from a ground level point source is:

D! = 0.457 EOX

X/Q= ruaya The surface body dose rate from beta emitters in the infinite cloud can be approximated as being one-half this Where amount (i.e., 0DD' = 0.23 E'X).

X = the short term average centerline value of the ground level concentration (curie/meter3)

For gamma emitting material the dose rate in air at the Q = amount of material released (curie/see)

u = windspeed (meter/see)

uloud center is:

y = the horizontal standard deviation of the plume (meters) [See Figure V-I. Page 48.

7 .D = 0.507 Ey(

Nuclear Safety, June 1961, Volume 2.

1.4-2

Number 4, "Use of Routine Meteorolo-ical Time Observations for Estimating Atmospcheric Following Dispersion," F. A. Gifford. Jrj.. Accident Atmospheric Conditions o" = the vertical standard deviation cf the pluii.e (meters) ISee Figure V-2, Page 48, Nuclear 0.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Pasquill Type F. wiudspeed I meter/sec.

Safqev', June 19(1. Volume 2. Number

4. uniform direction

"Use of Routlinc Me leorological Oh,'ervations for Estimating Atmospheric 8-24 hours Pasquill Type F, windspced I metcr/s.c.

Dispersion," F. A. G;ifford. Jr.I variable direction within a 22.5" sector

(.2) For lime periods of greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> the plume shouid hI assumed to meander and spread 1-4 days (a) 4(Y,,%( Pasquill Type D. windspeed 3 tini*ormly ovcr a 22.i" sector. The resultlant e'quaition is: rilel r/sec (b) 600,, Pasquill Type F. windspeed 2 leter/sec

2.032 x/Q = lx (W' wind direction v: riabie within a 22._.

sector

\Vhicrc

4-30 days (a) 33.35, Pasquill Type C, windspeed 3 x distance from point of release to the receptor;. meter/sec other variables are as given in g( 1). (N) 33.3%'. Pasquill Type D. windspeed 3 ineter/sec

(3) Tlhe at mospheric diffusion model" for (c) 33.3%; Pasquill Type F, wirdspeed 2 ground level releases is based on the information in the viieter/sec following lable. (d) Wind direction 33.3,:, frequency in a

22.50 sector

-This niIdo.l' %liould be useud until adequate site metcorologic'al d:ta are obtained. In some ,-uses. available information. such u,; tnic*orology. topography and geovaphicut. (4) Figures 2A and 213 give the groud level tocalion. may dictate Itic use of a more restrictive model to release atmospheric diffusion factors based on the insurc a conscrvative eltimuie of potentla oflfsitc exposures. parameters given in g( 3).

1.4-3

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