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 U.S. ATOMIC ENERGY COMMISSION

REGULATORY

DIRECTORATE

OF REGULATORY

STANDARDS Revision 1 June 1973 GUIDE REGULATORY

GUIDE 1.4 ASSUMPTIONS

USED FOR EVALUATING

THE POTENTIAL

RADIOLOGICAL

CONSEQUENCES

OF A LOSS OF COOLANT ACf',DENT

FOR PRESSURIZED

WATER REACTORS'

A. INTRODUCTION

Sect ion 50.34 o1f 10 CFR Pairl 50 requires that each applicant fir a c(nstruiction permit or operating license provid,: an analysis and cvalua3ion of the design and of structures.

systems, and components of tile facility with [he objective of assessing fhe risk to public health and safety resulting froim operation of the facility.

Tile design basis loss of" coolant accident (LOCA) is one of the postulated accidents Used to evaluate the adequacy of these structures, systems. and comiponents with respect to the public ltealth and safety.This guide gives acceptable assumptions that may be used in evaluating tIle radiologcal consequences of this accident for a pressurized water reactor. In some cases.unusual site characteristics, platit design features.

or other factors may require different assumptions which will be considered on an individual case basis. The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position.

B. DISCUSSION

After reviewing a number of applications for construction permits and operating licenses for pressurized wateli power reactors, the AEC Regulatory staff has developed a number of appropriately conservative assumptions, based on engineering judgment and on applicable experimental results from safety research programs conducted by the AEC and the nuclear industry, that are used to evaluate calculations of the radioloocal consequences of various postulated acciden ts.This guide lists acceptable assumptions that may be used to evaluate the design basis LOCA of a Pressurized Water Reactor (PWR). It should be shown that thc offsite dose consequences will be within thie guidelines of 10 CFR Part 100,'This guide is a revision of former Safety Guide 4.C. REGULATORY

POSITION 1. The assuimptions related io the release of radioactive material from the fuel and containment are as Ibllows: a. T we n t y -five percent of the equilibriut radioactive iodine inventory developed from imlaximu i full power operation of the core should be assumtned to be immediately available for leakage from the prinmary reactor containment.

Ninety-one percent of this 25 percent is to be assumed ito he ill Ithe forma ofelenllelllal iodine. 5 percent of this 25 percent ill the form of particulate iodine. and 4 percent of this 25 percent in the form of organic iodides.b. One hundred percent of the equilibrium radioactive noble gas inventory developed front maximum full power operation od the core should be assumed to be immediately available for leakage front the reactor containment.

c. The effects of radiological decay during holdup in the containment or other buildings should be taken into account.d. The reduction in the amotunt of radioactive material available for leakage to tile environment by containment sprays, recirculating filter systems, or other engineered safety features may be taken into account.but the amount of reduction in concentration of radioactive materials should be evaluated on an individual case basis.e. The primary reactor containment should be assumed to leak at the leak rate incorporated or to le incorporated as a technical specification requirement at 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 percent of this leak rate for the remaining duration of the accideint.

2 Peak accident pressure is the maximum1 pressure defined in the technical specifications for containment leak testing.2 Thte effect on coniainnmeni leakage tinder accident conditions of features provided to reduce the leakage ot" radioactive materials from the containment will be evaluated on an individual case basi

s. 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 the issuance or continuance of a permit or license by the Comrrssion.

1. Power Reactors 8. Products 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 and dose conversion are: a. The 0-8 hour ground level release concentrations may be reduced by a factor ranging from one to a maximum of three (.see Figure I) for additional dispersion produced by the turbulent wake of the reactor building in calculating potential exposures.

The volumetric building wake correction, as defined in section 3.3.5.2 of Meteorology and Atomic Energy 1968. should be used only in the 0-8 hour period: it is used with a shape factor of 112 and the minimum cross-sectional area of the reactor building only.b. No correction should be made for depletion of'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 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'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 the accident, the breathing rate should be assumed to be 1.75 x 104 cubic meters per second. After that until the end of the accident, the rate should be assumed to be 2.32 x 104 cubic meters per second. (These values were developed from the average daily breathing rate [2 x 107 cnv'/dayJ

assumed in the report of ICRP, Committee 11-1959.)d. The iodine dose conversion factors are given in ICRP Publication

2, Report of Committee

11,"Permissible Dose for Internal Radiation," 1959.e. 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" (Meteorology and Atomic Energy, Section 7.4. .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 in air at the cloud center is: From a semi-infinite cloud, the gamma dose rate in air is: ,D = 0,25E Where beta dose rate from an infinite cloud (rad/sec)gamma dose rate from an infinite cloud (rad/sec)EO3 = average beta energy per disintegration (Mev/dis)E = average gamma energy per disintegration (Mev/dis)X = concentration of beta or gamma emilling isotope in the cloud (curie/m3)

f. The following specific assumptions are acceptable with respect to the radioactive cloud dose calculations:

(1) The dose at any distance from the reactor should be calculated based on the maximum concentration in the plume at that distance taking into account specific meteorological, topographical, and other characteristics which may affect the maximum plume concentration.

These site related characteristics must 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 reactor. 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 always should be assumed to be at ground level.(2) The appropriate average beta and gamma energies emitted per disintegration, as given in the Table of Isotopes, Sixth Edition, by C. M. Lederer, J. M.Hollander, I. Perlman; University of California, Berkeley, Lawrence Radiation Laboratory;

should be used.g. The atmospheric diffusion model should be as follows: (1) The basic equation for atmospheric diffusion from a ground level point source is: X/Q= ruaya Where X = the short term average centerline value of the ground level concentration (curie/meter3)

Q = amount of material released (curie/see)

u = windspeed (meter/see)

y = the horizontal standard deviation of the plume (meters) [See Figure V-I. Page 48.Nuclear Safety, June 1961, Volume 2.D! = 0.457 EOX The surface body dose rate from beta emitters in the infinite cloud can be approximated as being one-half this amount (i.e., 0DD' = 0.23 E'X).For gamma emitting material the dose rate in air at the uloud center is: 7.D = 0.507 Ey(1.4-2 Number 4, "Use of Routine Meteorolo-ical Observations for Estimating Atmospcheric Dispersion," F. A. Gifford. Jrj..o" = the vertical standard deviation cf the pluii.e (meters) ISee Figure V-2, Page 48, Nuclear Safqev', June 19(1. Volume 2. Number 4."Use of Routlinc Me leorological Oh,'ervations for Estimating Atmospheric Dispersion," F. A. G;ifford.

Jr.I (.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 ovcr a 22.i" sector. The resultlant e'quaition is: 2.032 x/Q = lx\Vhicrc x distance from point of release to the receptor;.

other variables are as given in g( 1).(3) Tlhe at mospheric diffusion model" for ground level releases is based on the information in the following lable.-' 'This niIdo.l' %liould be useud until adequate site metcorologic'al d:ta are obtained.

In some ,-uses. available information.

such u,;

topography and geovaphicut.

tocalion.

may dictate Itic use of a more restrictive model to insurc a conscrvative eltimuie of potentla oflfsitc exposures.

Time Following Accident Atmospheric Conditions

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.

uniform direction 8-24 hours Pasquill Type F, windspced I metcr/s.c.

variable direction within a 22.5" sector 1-4 days (a) 4(Y,,%( Pasquill Type D.rilel r/sec (b) 600,, Pasquill Type F.leter/sec (W' wind direction v: riabie sector windspeed

3 windspeed

2 within a 22._.4-30 days (a) 33.35, Pasquill Type C, windspeed

3 meter/sec (N) 33.3%'. Pasquill Type D. windspeed

3 ineter/sec (c) 33.3%; Pasquill Type F, wirdspeed

2 viieter/sec (d) Wind direction

33.3,:, frequency in a 22.50 sector (4) Figures 2A and 213 give the groud level release atmospheric diffusion factors based on the parameters given in g( 3).1.4-3 bI .1-GiAK buPRIP IOýAO 2.5 FIGURE 1 O.SA-SO meters 2 0 P'mtr 2 4.0 .5A-1000 -nws O.SA-2500atr

2 meti 2 O.BA-1500

motors 0.5A-3000meru p. O.5A-2000

metonus ýw ccI 0 zI w1.5 4 I uII I I i ..I *10w Distance from Structure (metars)

I 3 L FIGL S1i.. ._10-A 7----4--I a -... *---- -- ......0 0\44sanc 4m St .tu. ... tes 1 o .* '1 %Ditne rmSrut- (ees I II _ .II I 1.4-5

10-I I I A ... ; I .... 1 1 I.", .---I.-- I--,- -.-*" i ... v -.1 -1 %ýFIGURE 2(81 -_____0-8 hours7~..,D L IU E 0.1.N.. I I ' I%, ...:.7 ::::: 'V t~rV~'I F-W 0 * ,p -I *.X1 ,-,I- --7I..~.i... I_... ... -10 163 1 1 .ý 10.5 b 7 1 Distance from Structure (maters)1.4-6