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{{Adams
{{Adams
| number = ML13350A195
| number = ML003739614
| issue date = 06/30/1973
| issue date = 06/30/1974
| title = Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors
| title = Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors
| author name =  
| author name =  
| author affiliation = US Atomic Energy Commission (AEC)
| author affiliation = NRC/RES
| addressee name =  
| addressee name =  
| addressee affiliation =  
| addressee affiliation =  
Line 10: Line 10:
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = RG-1.004, Rev. 1
| document report number = RG-1.4, Rev 2
| document type = Regulatory Guide
| document type = Regulatory Guide
| page count = 6
| page count = 6
}}
}}
{{#Wiki_filter:Revision 1 U.S. ATOMIC ENERGY COMMISSION
{{#Wiki_filter:Revision 2 June 1974 U.S. ATOMIC ENERGY COMMISSION  
REGULATORY
REGULATORY  
DIRECTORATE  
DIRECTORATE  
OF REGULATORY  
OF REGULATORY  
STANDARDS Revision 1 June 1973 GUIDE REGULATORY  
STANDARDS GUIDE REGULATORY  
GUIDE 1.4 ASSUMPTIONS  
GUIDE 1.4 ASSUMPTIONS  
USED FOR EVALUATING  
USED FOR EVALUATING  
THE POTENTIAL  
THE POTENTIAL  
RADIOLOGICAL  
RADIOLOGICAL  
CONSEQUENCES
CONSEQUENCES  
OF A LOSS OF COOLANT ACf',DENT
OF A LOSS OF COOLANT ACCIDENT FOR PRESSURIZED  
FOR PRESSURIZED  
WATER REACTORS  
WATER REACTORS'


==A. INTRODUCTION==
==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.
Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility.


systems, and components of tile facility with [he objective of assessing fhe risk to public health and safety resulting froim operation of the facility.
The design basis loss of coolant accident (LOCA) is one of the postulated accidents used to evaluate the adequacy of these structures, systems, and components with respect to the public health and safety. This guide gives acceptable assumptions that may be used in evaluating the radiological consequences of this accident for a pressurized water reactor. In some cases, unusual site characteristics, plant 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.
 
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==
==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
After reviewing a number of applications for construction permits and operating licenses for pressurized water 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 radiological consequences of various postulated accidents.
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.
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 the offsite dose consequences will be within the guidelines of 10 CFR Part 100. (During the construction permit review, guideline exposures of 20 rem whole body and 150 rem thyroid should be used rather than the values given in § 100.11 in order to allow for (a) uncertainties in final design details and meteorology or (b) new data and calculational techniques that might influence the final design of engineered safety features or the dose reduction factors allowed for these features.)
C. REGULATORY
POSITION 1. The assumptions related to the release of radioactive material from the fuel and containment are as follows: a. Twenty-five percent of the equilibrium radioactive iodine inventory developed from maximum full power operation of the core should be assumed to be immediately available for leakage from the primary 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 hours. and at 50 percent of this leak rate for the remaining duration of the accideint.
Ninety-one percent of this 25 percent is to be assumed to be in the form of elemental iodine, 5 percent of this 25 percent in the form of particulate iodine, and 4 percent of this 25 percent in the form of organic iodides.


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
b. One hundred percent of the equilibrium radioactive noble gas inventory developed from maximum full power operation of the core should be assumed to be immediately available for leakage from the reactor containment.


====s. USAEC REGULATORY ====
c. The effects of radiological decay during holdup in the containment or other buildings should be taken into account.
GUIDES Coples of published guldes may be obtained by request Indicating the divisions desired to the US. Atomic Energy Commission.


Washington.
d. The reduction in the amount of radioactive material available for leakage to the 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 be incorporated as a technical specification requirement at peak accident pressure for the first 24 hours, and at 50 percent of this leak rate for the remaining duration of USAEC REGULATORY
GUIDES Copies of published guide may. be obtained by request indicating the divisions desired to the US. Atomic Enemgy Washlngton.


0.1, 20545, Regulatory Guides are issued to describe and make avaliable to the public Attention:  
D.C. 20646, Regulatory Guides we issuad to describe and make available to the public Attention:  
Director of Regulatory Standards.
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.
Comments and suggestions for methods acceptable to the AEC Regulatory staff of implementing specific parts of Impr° ments In theose uldes we encouraged and should be sent to the Secretary the Commission's regulations, to delineate techniques used by the staff in of the Commislion, U.S. Atomic Energy Commission, Washington, D.C. 20645, eanluating specific problems or postulated accidents, or to provide guidance to Attention:
 
Chief, Public ProcoedlnglStaff.
O.C. 20545.evaluating specific problems or postulated accid3nts.


or to provide guidance to Attention:
applicants.
Chief, Public Proceedings Staff.applicants.


Regulatory Guides are not substitutes for regulations and compliance with them is not required.
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:
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 Comrrssion.
the guides will be acceptable if they provide a basis for the findings requisite to the Issuance or continuance of a permit or )iconse by the Commission.


1. Power Reactors 8. Products 2. Researcha nd Tast Reactors  
1. PeOWrd Reactors 6. Products 2. Research end Test Reactors  


===7. Transportation===
===7. Transportation ===
3. Fuels and Materials Facilities  
3. Fuels end Materials Facilities EL Occupatlonal Health Published guides will be revised periodically, as appropriate, to accommodate  
8. Occupational Health Published guides will be revised periodically, as appropriate, to accommodate  
4. Environmental and Siting 9. Antitrust Review comments and to reflect new information or experienca.
4. Environmental end Siting 9. Antlitrust Review comments and to reflect new informatio" or experience.


5. Materials and Plant Protection  
5. Materials and Plant Protection  
1
10. General the accident., Peak accident pressure is the maximum pressure defined in the technical specifications for containment leak testing.
 
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 1) for additional dispersion produced by the turbulent wake of the reactor building in calculating potential exposures.


===0. General ===
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 1/2 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.
.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 hours, the breathing rate of persons offsite should be assumed to be 3.47 x 10'cubic meters per second. From 8 to 24 hours 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
c. For the first 8 hours, the breathing rate of persons offsite should be assumed to be 3.47 x 10"4 cubic meters per second. From 8 to 24 hours 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 1.75 x 10-4 cubic meters per second. After that until the end of the accident, the rate should be assumed to be 2.32 x 10 4 cubic meters per second. (These values were developed from the average daily breathing rate [2 x 107 cm 3/day] assumed in the report of ICRP, Committee  
assumed in the report of ICRP, Committee 11-1959.)d. The iodine dose conversion factors are given in ICRP Publication  
11-1959.)  
2, Report of Committee  
d. The iodine dose conversion factors are given in ICRP Publication  
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  
2, Report of Committee II, "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).  
[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 7.4.1.1 -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)
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: SD4 = 0.457 fEX The effect on containment leakage under accident conditions of features provided to reduce the leakage of radioactive materials from the containment will be evaluated on an individual case basis.The surface body dose rate from beta emitters in the infinite cloud can be approximated as being one-half this amount (i.e., PD-1 = 0.23 Eox).  For gamma emitting material the dose rate in air at the cloud center is: ^/DL = 0.507 E&x From a semi-infinite cloud, the gamma dose rate in air is: 7 D = 0.25EYx Where 0 , = beta dose rate from an infinite cloudi(rad/sec)  
f. The following specific assumptions are acceptable with respect to the radioactive cloud dose calculations:
DI= gamma dose rate from an infinite cloud (rad/sec)  
E3 average beta energy per disintegration (Mev/dis)  
EF" = average gamma energy per disintegration (Mev/dis)  
X = concentration of beta or gamma emitting isotope in the cloud (curie/m 3) 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.
(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;  
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;
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)
Lawrence Radiation Laboratory;  
Q = amount of material released (curie/see)
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: 1 XIQ = u SrUayoz 1.4-2 Where X = the short term average centerline value of the ground level concentration (curie/meter
u = windspeed (meter/see)
3) Q = amount of material released (curie/sec)  
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.
u = windspeed (meter/sec)  
 
ay = the horizontal standard deviation of the plume (meters) [See Figure V-i, Page 48, Nuclear Safety, June 1961. Volume 2, Number 4, "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," F. A. Gifford, Jr.] z= the vertical standard deviation of the plume (meters) [See Figure V-2, Page 48, Nudear Safety, June 1961, Volume 2, Number 4, "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," F. A. Gifford, Jr.] (2) For time periods of greater than 8 hours the plume should be assumed to meander and spread uniformly over a 22.50 sector. The resultant equation is: 2.032 x/Q = uu OzU Where x = distance from point of release to the receptor;  
Jr.I (.2) For lime periods of greater than 8 hours 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(l). (3) The atmospheric diffusion model 2 for ground level releases is based on the information in the following table. 2 This model should be used until adequate site meteorological data are obtained.
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.
In some cases, available information, such as meteorology, topography and geographical location, may dictate the use of a more restrictive model to insure a conservative estimate of potential offsite exposures.


Time Following Accident Atmospheric Conditions
Time Following Accident Atmospheric Conditions
0.8 hours Pasquill Type F. wiudspeed I meter/sec.
0-8 hours Pasquill Type F, windspeed
1 meter/see, uniform direction
8.24 hours Pasquill Type F, windspeed
1 meter/sec, variable direction within a 22.50 sector 1-4 days (a) 40% Pasquill Type D, windspeed
3 meter/sec (b) 60% Pasquill Type F, windspeed
2 meter/sec (c) wind direction variable within a 22.50 sector 4-30 days (a) 33.3% Pasquill Type C, windspeed
3 meter/sec (b) 33.3% Pasquill meter/sec (c) 33.3% Pasquill meter/sec (d) Wind direction
22.50 sector Type D, windspeed
3 Type F, windspeed
2 33.3% frequency in a (4) Figures 2A and 2B give the ground level release atmospheric diffusion factors based on the parameters given in g(3).


uniform direction 8-24 hours Pasquill Type F, windspced I metcr/s.c.
==D. IMPLEMENTATION==
The revision to this guide (indicated by a line in the margin) reflects current Regulatory staff practice in the review of construction permit applications;
therefore, this revision is effective immediately.


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
1.4-3  
3 windspeed
0[-v _T_ ,-.44 -4 Rat -4-F-I H 1--r H:,ifif-----T -T 77 -- ---- --IW1ETýý T f-V 4 F I I FI I '-cli a'K-4 F I 'I-.T- 71* :1
2 within a 22._.4-30 days (a) 33.35, Pasquill Type C, windspeed
* I KY----9 -t-t -.."'III'ii F.V-~2 H 4~~~t 1-4 r~4 7.~I J:r __________________
3 meter/sec (N) 33.3%'. Pasquill Type D. windspeed
___414_____1___
3 ineter/sec (c) 33.3%; Pasquill Type F, wirdspeed
+1j -T4 LltIIII]I4J---I-1
2 viieter/sec (d) Wind direction
1V 11ff-14- -:4Ij4Th1 -i-i 0 W4 BUILDING WAKE DISPERSION
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
CORRECTION
2 meti 2 O.BA-1500
FACTOR 1'
motors 0.5A-3000meru p. O.5A-2000
* LrLii ilL-I U I I I..'-Iýff HT q---*--7- -7 7- -+~1*~7 17.......A .............Ii-w 7-----T 7-1-ý 4 Lt.]
metonus ýw ccI 0 zI w1.5 4 I uII I I i ..I *10w Distance from Structure (metars)
-i-v-ti-ti r-i 4 T I' I'.10-40 ::R+ 9 H-Lik i .F141 4 ERH ER HIE 1. 104 9 + L-f+ý+' A L -Lý V E t 4ý U. + #4-4ý tv ft 4 r ::1--, 4, 5 44-J 1-t -4 L-T-1, r -.4-.: '-4..,. '-44 -1-il 4". 1 2 -4 r' 1: I I tLT -I 4444 1-4 LV 4 4 rr Distance from Structure (meters)1.4-5 11111 iiiiiin::...,.
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
ri-Ill 74 19-4 9 6 4 3-t T, 44 444 RE-vT -T 1 S1 14 44 I -I 4 ft-6 I1Ilp -44 ý4, 1 10, FIGURE 2(AAMJ) i--T-4-+I-  
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}}
zh 4- 7ý+ý 4 T T' -T r! I L -4= T ý!J!ý 44-+ + 4-4-ý H V. r ..TT T! 9 l'o.F.U., 7777!ý i I li I I I I'llill'44
--... -.." I a I I I I! i ! ! I 17 I t44++H fffl -t-ýft ttlt ## 11E 4444 i i i i ! I-H fjP +4-tiHi++
ftHl" TTlq4lRh-  
4iý TfFffliif t 4. 4 H.4 -+j J+/- 44 -BE.1 A-'8 I I I
Disance fromt Structure (meter)1.4-6}}


{{RG-Nav}}
{{RG-Nav}}

Revision as of 18:16, 31 August 2018

Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors
ML003739614
Person / Time
Issue date: 06/30/1974
From:
Office of Nuclear Regulatory Research
To:
References
RG-1.4, Rev 2
Download: ML003739614 (6)
v • d • e


Revision 2 June 1974 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 ACCIDENT FOR PRESSURIZED

WATER REACTORS

A. INTRODUCTION

Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility.

The design basis loss of coolant accident (LOCA) is one of the postulated accidents used to evaluate the adequacy of these structures, systems, and components with respect to the public health and safety. This guide gives acceptable assumptions that may be used in evaluating the radiological consequences of this accident for a pressurized water reactor. In some cases, unusual site characteristics, plant 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 water 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 radiological consequences of various postulated accidents.

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 the offsite dose consequences will be within the guidelines of 10 CFR Part 100. (During the construction permit review, guideline exposures of 20 rem whole body and 150 rem thyroid should be used rather than the values given in § 100.11 in order to allow for (a) uncertainties in final design details and meteorology or (b) new data and calculational techniques that might influence the final design of engineered safety features or the dose reduction factors allowed for these features.)

C. REGULATORY

POSITION 1. The assumptions related to the release of radioactive material from the fuel and containment are as follows: a. Twenty-five percent of the equilibrium radioactive iodine inventory developed from maximum full power operation of the core should be assumed to be immediately available for leakage from the primary reactor containment.

Ninety-one percent of this 25 percent is to be assumed to be in the form of elemental iodine, 5 percent of this 25 percent in 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 from maximum full power operation of the core should be assumed to be immediately available for leakage from 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 amount of radioactive material available for leakage to the 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 be 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 USAEC REGULATORY

GUIDES Copies of published guide may. be obtained by request indicating the divisions desired to the US. Atomic Enemgy Washlngton.

D.C. 20646, Regulatory Guides we issuad 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 Impr° ments In theose uldes we encouraged and should be sent to the Secretary the Commission's regulations, to delineate techniques used by the staff in of the Commislion, U.S. Atomic Energy Commission, Washington, D.C. 20645, eanluating specific problems or postulated accidents, or to provide guidance to Attention:

Chief, Public ProcoedlnglStaff.

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 )iconse by the Commission.

1. PeOWrd Reactors 6. Products 2. Research end Test Reactors

7. Transportation

3. Fuels end Materials Facilities EL Occupatlonal Health Published guides will be revised periodically, as appropriate, to accommodate

4. Environmental and Siting 9. Antitrust Review comments and to reflect new information or experienca.

5. Materials and Plant Protection

10. General the accident., Peak accident pressure is the maximum pressure defined in the technical specifications for containment leak testing.

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 1) 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 1/2 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"4 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 1.75 x 10-4 cubic meters per second. After that until the end of the accident, the rate should be assumed to be 2.32 x 10 4 cubic meters per second. (These values were developed from the average daily breathing rate [2 x 107 cm 3/day] assumed in the report of ICRP, Committee

11-1959.)

d. The iodine dose conversion factors are given in ICRP Publication

2, Report of Committee II, "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 7.4.1.1 -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: SD4 = 0.457 fEX The effect on containment leakage under accident conditions of features provided to reduce the leakage of radioactive materials from the containment will be evaluated on an individual case basis.The surface body dose rate from beta emitters in the infinite cloud can be approximated as being one-half this amount (i.e., PD-1 = 0.23 Eox). For gamma emitting material the dose rate in air at the cloud center is: ^/DL = 0.507 E&x From a semi-infinite cloud, the gamma dose rate in air is: 7 D = 0.25EYx Where 0 , = beta dose rate from an infinite cloudi(rad/sec)

DI= gamma dose rate from an infinite cloud (rad/sec)

E3 average beta energy per disintegration (Mev/dis)

EF" = average gamma energy per disintegration (Mev/dis)

X = concentration of beta or gamma emitting isotope in the cloud (curie/m 3) 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: 1 XIQ = u SrUayoz 1.4-2 Where X = the short term average centerline value of the ground level concentration (curie/meter

3) Q = amount of material released (curie/sec)

u = windspeed (meter/sec)

ay = the horizontal standard deviation of the plume (meters) [See Figure V-i, Page 48, Nuclear Safety, June 1961. Volume 2, Number 4, "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," F. A. Gifford, Jr.] z= the vertical standard deviation of the plume (meters) [See Figure V-2, Page 48, Nudear Safety, June 1961, Volume 2, Number 4, "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," F. A. Gifford, Jr.] (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 plume should be assumed to meander and spread uniformly over a 22.50 sector. The resultant equation is: 2.032 x/Q = uu OzU Where x = distance from point of release to the receptor;

other variables are as given in g(l). (3) The atmospheric diffusion model 2 for ground level releases is based on the information in the following table. 2 This model should be used until adequate site meteorological data are obtained.

In some cases, available information, such as meteorology, topography and geographical location, may dictate the use of a more restrictive model to insure a conservative estimate of potential offsite exposures.

Time Following Accident Atmospheric Conditions

0-8 hours Pasquill Type F, windspeed

1 meter/see, uniform direction

8.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Pasquill Type F, windspeed

1 meter/sec, variable direction within a 22.50 sector 1-4 days (a) 40% Pasquill Type D, windspeed

3 meter/sec (b) 60% Pasquill Type F, windspeed

2 meter/sec (c) wind direction variable within a 22.50 sector 4-30 days (a) 33.3% Pasquill Type C, windspeed

3 meter/sec (b) 33.3% Pasquill meter/sec (c) 33.3% Pasquill meter/sec (d) Wind direction

22.50 sector Type D, windspeed

3 Type F, windspeed

2 33.3% frequency in a (4) Figures 2A and 2B give the ground level release atmospheric diffusion factors based on the parameters given in g(3).

D. IMPLEMENTATION

The revision to this guide (indicated by a line in the margin) reflects current Regulatory staff practice in the review of construction permit applications;

therefore, this revision is effective immediately.

1.4-3

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