ML19330A093

LER 80-095/03L-0:on 800617,during HNP-2-3123,loop Control Sys Blower Operability,Inboard Leadkage Control Sys Flow Indicator a Pegged Upscale & Would Not Return to Zero.Caused by Bent Flow Element Float.Float Straightened & Cleaned

ML19330A093
Person / Time
Site:	Hatch
Issue date:	07/07/1980
From:	Nix R GEORGIA POWER CO.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML19320C040	List: ML19320C039 ML19330A093
References
LER-80-095-03L, LER-80-95-3L, NUDOCS 8007150685
Download: ML19330A093 (2)

Text

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APPENDIX 15B i DOSE MODELS USED TO EVALUATE  !

THE ENVIRONMENTAL CONSEQUENCES ,

OF ACCIDENTS  ;

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PVNGS FSAR

.

APPENDIX 15B DOSE MODELS USED TO EVALUATE THE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15B.1 INTRODUCTION This section identifies the models used to calculate control room and offsite radiological doses, not calculated in CESSAR, that would result from releases of radioactivity due to various postulated accidents.

15B.2 ASSUMPTIONS The following assumptions are basic to the model for the whole body dose due to immersion in a cloud of radioactivity and to the model for the thyroid dose due to inhalation of radio-activity:

~ A. All radioactive releases are treated as ground level

{

[ releases regardless of the point of discharge.

B. The dose receptor is a standard man, as defined by the International Commission on Radiological Protec-tion (ICRP) ,. (reference 1) .

C. No credit is taken for cloud d,epletion by ground deposition and radioactive decay during transport to the exclusion area boundary (EAB) or the outer boundary qf'the low-population zone (LPZ).

D. Radionuclide data, including decay constants and decay energies presented in table 15B-1, are taken from references 2 through 6.

15B.3 WHOLE BODY GAMMA AND BETA SKIN DOSE The whole body gamma dose delivered to an offsite dose receptor is ca}cy% lated by assuming the receptor to be immersed T^ in a hemispherical radioactive cloud that is infinite in all 1

0302 -

elW4P 15B-1

PVNGS FSAR APPENDIX 15B Table 15B-1 RADIONUCLIDE PARAMETERS Average MeV/ Disintegration MeV/ Disintegration Nuclide Half-Life (gamma) (beta)

I-131 8.06 D 0.381 0.194 I-132 2.28 H 2.333 0.519 I-133 21 H 0.608 0.403 I-134 52 M 2.529 0.558 I-135 6.7 H 1.635 0.475 Kr-83m 1.86 H 0.002 0.037 Kr-85m 4.48 H 0.159 0.253 Kr-85 10.73 Y 0.002 0.251 Kr-87 76.31 M 0.793 1.324 Kr-88 2.80 H 1.950 0.375 Kr-89 3.16 M 1.712 1.001 Xe-131m 11.9 D 0.02 0.143 Xe-133m 2.25 D 0.0416 0.190 Xe-133 5.29 D 0.0454 .

0.135 Xe-135m 15.65 M 0.432 0.095 Xe-135 9.15 H 0.247 ,

0.316 Xe-137 3.83 M 0.194 l.642 Xe-138 14.17 M 1.183 '

O.606 l T.

directions above the ground plane; i . e . , 'a. emi-infinite cloud. The concentration of radioactive mat'drial within this 1 cloud is uniform and equal to the maximum centerline ground level concentration that would exist in the cloud at the appropriate distance from the point of release.

0303 ,

. g'

' ]

15B-2  ;

PVNGS FSAR APPENDIX 15B The gamma dose to an offsite receptor due to gamma radiation for a given time period is:

DCF

• Oi Dwb = X/O

• 1 wbi where D

wb

= whole body dose to an offsite receptor from gamma radiation, (rem) x/O = site atmospheric dispersion factor effective during the time period at the point of exposure, 3

(s/m )

DCF whole body dose conversion factor for the semi-bi =

infinite cloud model for nuclide i, (rem-m 3 /Ci-s).

(See table 15B-2)

Q. = total activity of nuclide i released during the j time period, (Ci)

V The gamma dose to the control room personnel is calculated assuming a finite hemispherical cloud model. The gamma dose due to gamma radiation in the control room for a given time period is:

D _ (CRV'O$) 0. 338 DCF wb1

( i} (

wb ~ 1173 i (CRVOL) ( 0. 02832 )

where D

wb

= whole body gamma dose to control room personnel from gamma radiation, (rem)

CRO = the control room occupancy factor $1 3600 = conversion factor, s/h

.02832 = conversion factor, ft /m CRVOL = control room volume, ft

,.:. 4 :n '~0304 i g 15B-3

PVNGS FSAR APPENDIX 15B Table 15B-2 WHOLE BODY GAMMA AND BETA SKIN DOSE CONVERSION FACTORS Beta Skin DCF Whole Body Gamma DCF Radionuclide (rem - m3/Ci - h) (rem - m3/Ci - s)

'I-131' l.14E2 8.72E-2 I-132 4.75E2 5.13E-1 I-133 2.65E2 1.55E-1 I-134 3.32E2 5.32E-1 I-135 4.64E2 4.21E-1 Kr-83m 0 5.02E-6 Kr-85 1.53E2 5.25E-4 Kr-85m 1.67E2 3.72E-2 Kr-87 1,llE3 1.87E-1 Kr-88 2.70E2 4.64E-1 Kr-89 1.15E3 5.25E-1 Xe-131m 5.43El 2.92E-3 Xe-133m 1.13E2 8.00E-3 lh Xe-133 3.49El 9.33E-3 Xe-135m 8.llEl 9.92E-2 Xe-135 2.12E2 5.72E-2 Xe-137 1.39E3 4.53E-2 Xe-138 4.71E2 2.81E-1 1

10 1 = total integrated activity for nuclide i in control room for the time period, (Ci-hr)

DCF the semi-infinite cloud whole body gamma dose wbi =

conversion factor for nuclide i, (rem-m 3/Ci-s).

(See table 15B-2)

• 0 The expression (CR O is a geometrical correction factor to ratio a finite cloud to infinite cloud (reference 7).

O oaos 9 15B-4

f PVNGS FSAR APPENDIX 15B 0(./ The beta skin dose to control room personnel is calculated .

assuming a tissue depth of 7 mg/cm2 . The beta skin dose to control room personnel for a given time period is:

D ED -

Of (3)

Ss (CRVOL 02832) 1 Ssi where D =

the beta skin dose conversion factor for nuclide i, 6si 3

(rem-m /Ci-h). (See table 15B-2 for factor) and all other parameters are as previously defined.  ;

15B.4 THYROID INHALATION DOSE The thyroid dose to an offsite receptor for a given time period is obtained from the following expression:

D = X/Q B I ( -

DCFf) (4) x where:

D = thyroid -inhalation dose, (rem)

X/0 =

site. atmospheric dispersion factor during the time period, (s/m 3)

B =

breathing rate during the time period, (m /s)

, (See table 15B-3) 01 =

total activity of nuclide i released during time period, (Ci)

DCF =

thyroid dose conversion factor for nuclide i, i

(rem /Ci inhaled). (See table 15B-4)

The radionuclide data are given in table 15B-1. The atmospheric dispersion factors used in the analysis of the environmental consequences of accidents are given in section 2.3.

Y&

o 6

15B-5

PVNGS FSAR APPENDIX 15B Breathing rates and dose conversion factors for radioactive iodines required for computing thyro'id inhalation doses are tabulated in tables 15B-3 and 15B-4, respectively.

Table 15B-3 BREATHING RATES ("}

Time After Accident m /s 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.47(-04) 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.75(-04) 1 to 30 days 2.32(-04)

a. From Regulatory Guide 1.4 Table 15B-4 IODINE DOSE CONVERSION FACTORS ("}

Iodine Isotope (rem-thyroid / Curie Inhaled) ,

I-131 1. 4 8 (+0 6)

I-132 5. 3 5 (+0 4 ),

I-133 4.00(+05)

I-134 2. 5 0 (+ 0,4 )

I-135 1. 2 4 (+05 )~

a. See reference 8 ISB.5 CONTROL ROOM DOSE

• During the course of an accident, control room personnel may receive doses from the following sources: "

A. Direct whole body gamma dose from the radioa,ctivity present in the containment building ,

B.. Direct whole body gamma dose from the radioactive

'~, cloud surrounding the control room 0307 M.

15B-6  !

PVNGS FSAR APPENDIX 15B

(

C. Whole body gamma, thyroid inhalation, and beta skin

( )

doses from the airborne radioactivity present in the control room.

In calculating the exposure to control room personnel, occupancy factors were obtained from reference 7 as follows:

0 to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: occupancy factor = 1 1 to 4 days: occupancy factor = 0.6 4 to 30 days: occupancy factor = 0.4 The dose model for each of the radiation sources are discussed below:

A. Direct whole body gamma dose from the radioactivity present in the containment building (direct containment dose).

Time integrated (0 to 30 days) radionuclide concentra-

,y

- tions in the containment are calculated. For

(_) conservati< redit is taken for reduction of the

.containmen- ..etivity by means other than radioactive decay. The containment is modeled by an equivalent volume cylindrical source having a diameter of 146 feet and a height of 155 feet. The radioactivity present in the containment is assumed to be uniformly distributed in the cylindrical source. Shielding is provided by the 4-foot concrete containment walls, 120 feet of air separating the containment building

,from the control building, and 2-foot thick control room walls.

No credit is taken for any shielding that would be provided by the auxiliary building.

B .- Direct whole body gamma dose from the radioactive cloud surrounding the control room (outside cloud dose) .

y 0308

~

15B-7

PVNGS FSAR APPENDIX 15B Leakage from the containment building, or any building h

~

will result in the formation of a radioactive plume.

For conservatism it is assumed that this plume forms a cloud surrounding the control room. Gamma radiation from this cloud, although attenuated, can penetrate the control room roof and walls resulting in a whole body gamma dose to control room personnel. The radius of the cloud is computed using a mass balance of the radioactivity released due to leakage and the volume of the cloudt therefore, the radioactive cloud is time variant and expands for the duration of the accident.

Radioactivity concentration (Ci/m ) in the radio-active cloud surrounding the control room is the product of the building leak rato (Ci/s) and the control room atmospheric dispersion factor, X/0 (s/m 3),

Exclusion area boundary and low population zone X/Q's are presented in section 2.3. A tabulation of control room X/Q's is presented in table 15B-5.

The calculational model for the r;.untrol room is an equivalent volume hemisphere of radius 42 feet. Credit is taken for concrete shielding provided by the control room walls and ceiling.

Table 15B-5 ATMOSPHERIC DISPERSION FACTORS Time Period Control Room X/Q (s/m )

0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.97(-3) 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.16(-3) 1 to 4 days 4.53(-4) 4 to 30 days 1. 3 0 (~4 )

O 15B-8

PVNGS FSAR

- APPENDIX 15B

\ '

/ C. Dose from the airborne radioactivity present in the control room (occupancy dose).

Airborne radioactivity will be drawn into the control room due to the intake of outside air required to maintain a positive pressure in the control room.

This contributes to the whole body gamma, thyroid inhalation, and beta skin doses. The major parameters of the control room ventilation system are presented in table 15B-6.

The whole body gamma dose is computed using a finite cloud model. The calculational model is an equivalent volume hemisphere of 42-foot radius.

A thyroid inhalation dose results from the radioactive iodine present in the control room. The control room habitability system, designed to renave iodine from the air, is described in table 15E-6.

b('N 15B.6 . ACTIVITY RELEASE MODELS 15B.6.1 GENERAL EQUATION The activity released from a postulated accident is calculated by using the following matrix equation for each isotope and each specie of iodine:

hf+cX=S; Initial Condition K(t )g = K o (5)

Q=L - AI where:

K(t) =

(a1(t))

ag = the activity in the ith node, (Ci)

C = (C matrix

-* 1 >:

i

.'g)

( )

</

2\9 15B-9

PVNGS FSAR APPENDIX 15B Table 15B-6 ,

CONTROL ROOM ESSENTIAL VENTILATION SYSTEM PARAMETERS ("

Parameter Assumption Number of emergency ventilation systems 1 operating Filtered Intake rate, standard ft / min 1,000 Unfiltered intake rate, standard ft / min 0 Intake cleanup filter efficiency Iodine, elemental, % 95 Iodine, organic, % 95 Iodine, particulate, % 99 Recirculation rate, standard ft / min 27,410 Recirculation cleanup filter efficiency Iodine, elemental, % 95 Iodine, organic, % 95 Iodine, particulate % 99 3

Leak rate, standard ft / min (out leakage) 1,000 Control room volume, standard ft 161,000

a. There are two completely redundant emergency control room ventilation systems.

For a more detailed description of this system, refer to section 9.4.2. The dose model employed in this analysis is consistent with the thyroid inhalation model discussed in section 15B.4.

The beta skin dose model is consistent with the ,

" infinite hemispherical cloud" model described in  ;

section 15B.3.

l l

0311 g

15B-10 1

PVNGS FSAR APPENDIX 15B

) C g = the transfer rate from the ith noJe to the jth node, (s-1)

(S1) vector S =

Sg = the production rate in the ith node (Ci/sec)

Q = the activity released to the environment over the time period t g to t g, (Ci) 5 =

(ty) matrix 1

1

= the leak rate from the ith node to environment

( /sec)

E =

A(t) dt (Ci-sea) o Each node represents a volume where activity can be accumulated.

The environment and the control room are each represented by a node. To ensure that the system of differential equations has constant coefficients, the time scale is broken up into time intervals over which all parameters are constant. Thus, all coefficients and sources are assumed to be representable by step functions.

The matrix equation is solved using matrix techniques. The particular solution is obtained by Gaussian elimination. The homogenous solution is obtcined by solving for the eigenvectors and the eigenvalues of the coefficient matrix C. They are determined by using QR transformation techniques.

The following sections describe how the coefficient matrix and the source vector are calculated for the different accident calculations.

0312 15B-ll l

PVNGS FSAR APPENDIX 15B 15B.6.2 THE MODEL FOR CONTAINMENT LEAKAGE The model for LOCA containment leakage is shown in figure 15B-1.

The system of differential equations for estimating the released activity is as follows:

dA y+A A (6a) dt d1 -b21^2 -b31^3 =0 dA dt + ( d+ s +b21 + L23)^2 -b32^3 =0 (6M dA dt

-b 2 3' 2 + I d+L31 + b32)^3 = 0 (6c) dA dt Q

Ib u * (1 L f b

21 ^2 -h u+ IL b f 31 ^3 (6d)

+ (Lg+L +fRpc+ d} ^4 =0 t

O= (L21 ^2

• b31 A) 3 dt (7) t where:

Ay(t) = activity in the environment, (Ci)

A2(t)

= ctivity in the sprayed region of the containment, (Ci)

= activity in the unsprayed region of the contain-A3 ( t) ment, (Ci)

A4(t)

= activity in the control room, (Ci)

A d

= radioactive decay constant, (s-l)

T 21 b

21 (100)(24)(3600), (s-1)

T = leak rate from the sprayed volume to the environ-21 ment (%/ day)

T b

31 31 (100)(24)(3600), s -l)

T = leak rate from the unsprayed volume to the environ-31 ment (%/ day) ,

5B-12 0313 Y L

PVNGS FSAR APPENDIX 15B

-[ A s

= th spray removal constant, (s-1)

L 23 "

(v ) (60) , (s )

2 T = transfer rate from the sprayed region to the 23 3

unsprayed region, (ft / min) 3 V = volume of the sprayed region, (ft )

2 T

32 -1 32 (v 3) (60) ' U I T = transfer rate from the unsprayed region to the 32 sprayed region, (ft / min) 3 V = v lume f the unsprayed region, (ft )

3 T -

(.3048)3 L

u 60 , (m /s)

T = unfiltered inleakage into the control room, u

(ft /3 min)

T (.3048)3 Lg =

60 , ( /sec)

T = filtered air intake rate into the control room, f

(ft /3 min) fg = filter efficiency of the filters on the intake units a

X/O = atmospheric dispersion factor for the control room, (s/m 3)

T r

R c

(V ) (60) ,

s -1) c T = filtered recirculatior, rate in the control room, R

(ft /3 min)

V = control room free. volume, (ft 3) c f

3

= filter efficiency of the filter on the recircula-tion unit Q = activity released to the environment, (Ci) a u,.,,

a s '0314 15B-13

PVNGS FSAR APPENDIX 15B The coefficient matrix is:

C=

+A -b -b 0 d 21 31 0 0

+(Ad+ s+b21+b23 - 32 0 -L 0 23 *I d+b31+b32) 0

-h(L+II-fILIb u L f 21 -hIb+II-E) u L bib 31 f

+( f+bu+fR c+ d}

Af ter solving for A(t) , the integrated activity in each node can then be calculated.

From the integrated activity, the offsite doses and the doses to the operators in the control room can be calculated using the dose models given in sections 15B.3 and 15B.4.

15B.6.3 THE MODEL FOR RECIRCULATION LOOP LEAKAGE The model for LOCA leakage in recirculation loops outside containment is shown in figure 15.B-2. The activity released due to the operational leakage of the engineered safety feature (ESF) components during the recirculation mode of the l postulated LOCA is calculated from the following equations:

dA dt

+

dlA - Il-II b 21 A 2

-0 (8a) dA

+ *8 dt (+ d+L21) A 2 2 t

1 Q= (1-f) L 21 A 2 dt (9) tg where:

A1 = the activity in the environment, (Ci)*

A2 = the activity in the ESF component rooms, (Ci) 0315 15B-14

PVNGS:FSAR APPENDIX 15B

/'

Ad = decay constant, (s~)

L21 = filtered leak rate to the environment, (ESF room vol/s) ,

f = filter efficiency of the filters on the ESF room purge i

units.

i A T U 8

S2=P -

y s

Ag = activity in the recirculation water, (Ci)

P = iodine partition factor 3

T = twice the maximum operational leak rate, (cm /s) s

! 3 V g = total volume of recirculation water, (cm )

Q = activity released to the environment, (Ci)

! The coefficient matrix is:

"A -(1-f)L 21 d

C=

0 (A d+L21)

, +

I The source vector is S=

8

L 2_

j 15B.6.4 THE MODEL FOR THE FUEL HANDLING ACCIJENT IN THE FUEL BUILDING WITH ESF SAFEGUARDS ACTUAT'.ON The model for the release of activity from **.te fuel building during a postulated fuel handling accident is shown in I' figure 15B-3. The activity released to the environment is l estimated-from the following equations:

.dA y+AA dt- dl- Il~f) L 21A 2 = 0 , (10a) i

. (J c1Gu

(

9

-15B ,. - - , . _ . - . . .. : - .---~. ._ .... - . _ - - - .

PVMGS FSAR APPENDIX 15B dA dt 2+ I d+b21)^2 = 0 (10b) ty  ;

O= L A dt 21 2 (11) o where:

Ay = activity in the environment, (Ci)

A = a tivity in the fuel building atmosphere, (Ci) 2 A

d

= decay constant, (s-1)

L 21 = purg rate to the environment, (s-1) f = filter efficiency of the filters on the ventilation unit Q = activity released to the environment, (Ci)

The resultant coefficient matrix is:

"A.d - (1-f) L 21 O

(Ad+ 21}

15B.6.5 OTHER ACCIDENT MODELS Other accidents can be conservatively modeled as simulated instantaneous releases to the environment. This is simulated as a large transfer rate to the environment. The model is shown in figure 15B-3. The system of differential equations is:

dA y+A u dt dl -b21^2 =0 (12a) dA dt 2* IAd+b21) ^2 =0 (12b) fty L (13)

Q=J 21 ^2 dt o

0317 O 15B-16

PVNGS FSAR APPENDIX 15B I  :

1/ where:

Ay = activity in the environment, (Ci)

A = activity to be released to the envitoninent, (Ci) 2 A

d

= decay constant, (s-1)

L21 = very large transfer rate to the environment, (s-1)

Q = activity released to the environment, (Ci)

The resultant coefficient matrix is:

"A -b d 21 C=

0 (Ad+ 21'_

15B.7 REFERENCES

1. " Report of ICRP Committee II, Permissible Dose for Internal

(}

'^'

Radiation (1959)," Health Physics, 3, p 30, 146-153, 1960.

2. Martin, M. J. and Blichert-Toft, P. H., Radioactive Atoms, Auger-Electron, L, B, y, and X-Ray Data, Nuclear Data Tables A8, 1, 1970.

3. Martin, M. J., Radioactive Atoms - Supplement 1, ORNL-4923, August 1973.

4. Bowman, W. W. and MacMurdo, K. W., " Radioactive Decay Gammas, Ordered by Energy and Nuclide," Atomic Data and Nuclear Data Tables 13, 89, 1974.

5. Meek, M. E. and Gilbert, R. S. , " Summary of Gamma and Beta Energy and Intensity Data /' NEDO-12037, January 1970.

6. Lederer, C. M., Hollander, J. M., and Perlman, I., Table of the Isotopes, 6th edition, March 1968.

7. Murphy, K. G. and Campe, K. M., " Nuclear Power Plant Control Room Ventilation System Design for Meeting General

.f ^' Criterion 19," Thirteenth AEC Air Cleaning Conference.

D318 -

M 15B-17

x PVNGS FSAR APPENDIX 15B

8. Di Nunno, J. J., et. al., " Calculation of Distance Factors for Power and Test Reacter Sites," TID 14844, March, 1962.

O I

I

.. i e g a ; . '

0 3,ai a 15B-18

(3 CONTAINMENT UPSPR AYED REGION L21 V

32 J L '23 L CONTROL ATMOSPHER E V ROOM

'31 V

SPRAYED REGION O

Q DIRECT UNFILTERED LE AKAGE FR ACTION FROM v UNSPR AYED R EGION Q DIRECT UNFILTERED LEAKAGE FRACTION FROM v SPRAYED REGION

.c TR ANSFER R ATE FROM SPRAYED REGION TO UNSPR AYED REGION TR ANSFER R ATE FROM UNSPR AYED REGION TO SPRAYED REGION Palo Verde Nuclear Generating Station FSAR I ,

, ( s' CONTAINMENT LEA E I

DOSE MODEL 04'O Figure 15B-1 r, g

..-. - _ . . = . . . .. _ _ . _ . _ - . . . - . - - . _ - . . -. __ . . . . . - ~ .. - . _ . - . - . _ . . - . .. .

i

! l

. f i

1 i

{

1 i

i i

RO MS ATMOSPHERE I '

r r FILTER -

r SOURCE i

1 i

i i

i l

• Q RECIRCULATION OF SUMP WATER TO ESF  !

4 .

v OOMPON ENTS '

h DIRECT FILTERED LEAKAGE FRACTION FROM ESF ROOM i

i f

I-Palo Verde Nuclear Generating Station FSAR l ESF ROOM LEAKAGE DOSE MODELS

- 0321 Figure 15B-2

^

h s

e si m v

PR IM ARY HOLDUP ATMOSPHERE SYSTEM y FILTER y (G

,l v

Q DIRECT UNFILTERED LEAKAGE FRACTION FROM V PP' MARY HOLD-UP SYSTEM Q DIRECT FILTERED LEAKAGE FR ACTION FROM v PRIMARY HOLDUP SYSTEM l

Palo Verde Nuclear Generatistg Station l

C g

0322 OTHER ACCIDENT DOSE MODEL Figure 15B-3 1

Jaur, I