ML20134M767
| ML20134M767 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 08/31/1985 |
| From: | Michael Cheok, Umbarger J NUS CORP. |
| To: | |
| Shared Package | |
| ML20134M750 | List: |
| References | |
| NUS-4758, NUDOCS 8509040248 | |
| Download: ML20134M767 (34) | |
Text
. . . . . - . . . - - .- - _
- +
1 i
d NUS 4758
.i 1
CONTROL ROOM RADIOLOGICAL REANALYSIS BRUNSWICK STEAM ELECTRIC PLANT-i i
1 Prepared for CAROLINA POWER & LIGHT COMPANY By M. C. Cheok i
i i
i August 1985 I' -
- . < .]. i.
- f j l Approved by: 8/ u %47 M//
Umbarger, Manager
.fAnalyticalvServices
.A. Department Consulting Division 4
b NUS-CORPORATION
.910 Clopper Road
- Gaithersburg, Maryland 20878 t
1 l
8509040248 050830 PDR F ADOCK 05000324 PDR i - - - . - - . . , - . . . - , - - , . , . - . , . - . , . .-.
4 TABLE OF CONTENTS Section and Title Page
1.0 INTRODUCTION
....................................... 1-1 2.0 METHODS............................................ 1-4 3.0 ASSUMPTIONS........................................ 1-4 4.0 RESULTS............................................ 1-7
5.0 REFERENCES
......................................... 1-11 8
4, f
i 1
i
! i I
' NUS CORPOAATION
J 1
1.0 -INTRODUCTION 1-This report summarizes the methods and results of the analysis of control room habitability during postulated radiological j
accidents at the Brunswick Steam Electric Plant. The analysis is performed in response to NUREG-0660, Itdm III.D.3.4 " Control lj Room Habitability" which requires licensees to ensure that plant t
- - operators are adequately protected from radiation and other hazards and that the control room can be used in the event of an f emergency.
l A previous report by NUS (NUS-3697, 1983) addressed the occupational doses (whole body, thyroid, and beta) to personnel j located inside the control room at 30 days following a l4 postulated loss-of-coolant accident (LOCA). This analysis was based on the simplified HVAC flow diagram shown in Figure 1.
The analysis considered the dose contribution inside the control j room from airborne radioactivity released through_ the plant
) stack as well as the containment shine and direct dose (gamma, whole body) from other sources shown in Table 1. The results
- shown in Table 1 demonstrated that the doses would not exceed General Design Criteria 19 limits.
1 This report considers the dose impact of increasing the unfiltered inleakage direct into the control room to 3000 cf:r along with the following four combinations of recirculated flow
- and flow through the charcoal filter:
I
- Filtered Intake Recirculated Flow 4
- (cfm) (c h)
Case 1 1000 1000
. Case 2 1500 500 Case 3 3000 500 Case 4 2500 1000 i
1-1
- _ NUS COAPOAATION
i 1,
100 cf m 998cfm 37.878 cfm A
i j
3E w-O Control Room 4cfm n e"
_ t i Units 1 and 2 2 l 5 u y <C ( - o o > 40.000 +
998cfm I3 4
? :E
? } ctm
?
898cfm
=0 I i Er m
14cfm Total Volume 83 efm
= 293.650 ft3 db NOTES y Total inieaxage tnat evoasses enarcoal fitte* 1374 cfm 150cfm
= 4 + 14 - 20 + 5 + S 3 - 150
= 276 c'm Totai enseanage tnat goes throu;n enarcoal filters = 100 cfm t
Source: Reference 2 P
l Figure 1. Brunswick Control Room HVAC System Flow Diagram 1-2 NUS COAPOAATCN
TABLE 1 BASELINE RESULTS OF RADIOLOGICAL ANALYSIS OF THE BRUNSWICK CONTROL ROOM i
Dose (rem)
'j Source of Radiation Whole Body Thyroid Beta Skin
! Airborne radioactivity released from the SGTSa through the plant stack 0.002 0.41 0.039 Reactor building--shielded portion 0.004 -- --
Reactor building--refueling area 0.109 -- --
SGTS charcoal filters 0.001 -- --
Control room charcoal filter 0.054 -- --
Cloud outside the control room 0.160 -- --
Core spray line within the reactor building 0.085 -- --
Stack structure shine Negligible -- --
Total 0.415 0.41 0.039 astandoy Gas Treatment system.
f I
1-3 NUS COAPCAATION
b Additionally, the maximum permissible unfiltered inleakage
~
l directly into the control room is determined for the above cases i
j by setting the dose at General Design Criteria 19 levels (see
{ - also Standard Review Plan Section 6.4), that is ,
- Whole body - 5 rem Thyroid - 30 rem Beta skin - 30 rem 2.0 METHODS The methods used _ to calculate the beta and gamma whole body
' doses and the thyroid dose to the control room operators- are standard calculational techniques for modeling the generation, release, . transport, buildup, and removal of radionuclides. The equations used to model these phenomena are incorporated into I
j the computer program used in this study to calculate the control i room operator doses and are presented in Appendix A of this report. The methods used to compute the whole body dose f
1 contributed by sources of direct radiation outside the control room are based on the work of Jaeger, Chapter 6 (Ref. 4). The i shine dose from liquid source terms was present;ed in Carolina Power and Light Company's response to NUREG-0578 Item 2.1.6.b.
1 l 3.0 ASSUMPTIONS io i The assumptions used in this analysis of control room radiation
- l exposures are described below and in Table 2:
1!
l.
- . o Radionuclides released from the reactor core are uni-
- ! formly distributed throughout the primary containment.
I '
Radionuclides released to the secondary containmen:
are assurred . to be uniformly distributed ~ throughout the j secondary containment.
f 4
1-4 NUS COAPOA ATION
- e TABLE 2 ASSUMPTIONS IN RADIOLOGICAL ANALYSIS OF TliE BRUNSWICh CONTROL ROOM Power level = 2,550 MWt Operating time = 1,000 days Fraction of core radionuclide inventory released to drywell Noble gases = 100 percent Halogens = 25 percent i
Drywell free volume = 164,000 ft3 Maximum / minimum wetwell free volume = 134,600/124,000 cfm Reactor building free volume = 2,000,000 ft3 Standby gas treatment system flow rate = 3,000 cfm Standby gas treatment system filter efficiencies for iodine Elemental = 95 percent Organic = 95 percent Particulate = 99 percent Primary containment leak rate = 0.5 percent / day Secondary containment air exchange rate = 100 percent / day Control room volume = 298,650 f 3 Control room ventilation system filter efficiencies for iodine Elemental = 95 percent Organic = 90 percent Particulate = 95 percent Stack height = 100 meters Atmospheric diffusion factors for stack release Time X/Q value (sec/m3) 0-1/2 hr 3.3 x 10-4 1/2-8 hr 1.8 x 10-6 i
, 8-24 hr 1.1 x 10-6 i 1-4 days 2.0 x 10-7 1 4-30 days 2.7 x 10-8 '
1-5 l
1 NUS COAPOAATION
o The primary containment leaks at a constant rate of 0.5 percent per day for the duration of the accident.
o The primary containment is assumed to consist of a single volume with no washout of radionuclides by con-tainment spray.
I o The secondary containment exhaust rate is assumed to l be one secondary containment volume per day.
I o Tnere is no direct leakage from primary containment to the environment.
o Based on information provided in the Brunswick FSAR, and the recommendations of USNRC Regulatory Guide 1. 3 ,
all exhaust from the secondary containment to the environment is assumed to pass through the standby gas treatment system. (Tne containment isolation system is designed to maintain a negative pressure of 0.25 inches of water in the secondary containment structure thus preventing leakage direct into the environment.
It also provides a means for minimizing the release of radioactive ' material to the environment by filtering the radioactivity and tnen exhausting it enrough a stack as an elevated plume.)
o The dose calculation does not include consideration of MSIV leakage.
o Tne post-accident core activity inventories are based on a continuous full power operation of 2550 MWt for 1000 days. These activities are based on TID-14844 using the fractional iodine release assumptions of USNRC Regulatory Guide 1.3.
, 1-6 NUS COAPOAATION
p .
+
o The accident duration is assumed to be 30 days.
- o Radionuclides in the control room are assumed to be
, uniformly distributed throughout that volume.
l o The breathing rate of the control room operators is
- ! - assumed to be 3'47 x 10-4 cubic meters per second for i
the duration of the accident.
i o The control room X/Q values are adjusted for the
{
l occupancy factors given in NRC Standard Review Plan 6.4. l*
'* 4.0 RESULTS i
11 i The radiation dose to individuals within the control room during a postulated design basis accident at the Brunswick station is
- computed using the assumptions above and those presented in Table 2 and in Appendix A. The meteorological data and HVAC design parameters are based on the information presented in Sec-t j tions 3.0 and 4.0 of Reference 2.
1 i, As described in the Brunswick FSAR, the maximum calculated dose
! to individuals within the control room occurs during a postu-l, lated LOCA. This is because the' magnitude and duration of the
. radionuclide release during a LOCA is much greater than that for any other accident. This is discussed further in References 5 and 6. I 1 The dose to control room personnel from radioactivity buildup-j within the control room for the four cases of interest is cal-
- culated using the HVAC system model and the data shown in l Figure 2. This figure shows the possible inleakage paths into
! the ductwork and into the control room itself. The 30-day inte-l grated dose caused by airborne radioactivity within the control
- l i
, 1-7 d ! NUS COAFOAATION i _. , _ , . . , _ . , . . . _ _ . , - - , . _ . . _ _ . , _ _ _ , _ , , . _ ~ _ . . . . . . ~ . . . _ _ - _ - . . _ _ . _ , _ , , .
100cfm B cfm o 4 1 37.878 cfm A 1f
=
i5 ej Control Room 1 _ 4 cfm -E E Units 1 and 2 1 2 l w 1 1r <c 4 w i, i, > E A cfm : i5 ? ? ? +
=0 =g " ctm D cfm l f
- Et g 14 cfm Total volume 83 cfm
= 298.650 ft3 ?
4> if F cfm C cfm NOTCS: Flow Rates (cfm) A B C D E F Case 1 1000 1000 3000 900 40,004 4226 Case 2 1500 500 3000 400 40,004 4726 Case 3 3000 500 3000 400 41,504 6226
- Case 4 2500 1000 3000 900 41,504 5726 1
f Figure 2. Brunswick Control Room HVAC System Flow Diagram--Sensitivity on Various Filtered and Unfiltered Flow Rates 1-8 1 ii i NUS CORACA ATION
room is summarized in Table 1 for the " baseline" case considered in Figure 1. Also provided are the gamma or whole body doses from various sources located outside of the control room, including containment shine. The dose contribution from these other sources is estimated to be 0.413 rem and is not affected by the changes in HVAC flow considered in this study. Table 3 summarizes the results of this study, presenting the airborne and total expected occupational doses for the additional four ventilation system cases considered in the study. The dose due to reactor building shine was calculated using the l QAD computer code (described in Appendix A). The features of the reactor and control building essential to the control room i shielding analysis were input in the QAD code. The sources used
! in the QAD code were based on the method described in Appendix A. These results are based on a reactor building concrete wall thickness of 2.0 feet and control building wall and roof concrete thicknesses of 2.0 feet each.
As shown in Table 3, the calculated control room doses are well within the current NRC criteria of 5 rem whole body, 30 rem thyroid, and 30 rem beta skin for all four cases. As discussed, a sensitivity study was performed to determine the maximum permissible unfiltered inleakage into the control roor
, by setting the dose levels at the above mentioned NRC criteria.
This study shows that the doses from airborne radioactivity in >l the control room will peak and level off at 2.6 rem thyroid, 0.004 rem whole body and 0.06 rem beta skin for unfiltered inleakage of 100,000 cfm or greater. At this level cf inleakage, the doses are limited by the source term (i.e., the amount of radioactivity released from the secondary containment; and are independent of the control room flow parameters. l-9 NUS COAPOAATION
i . a TABLE 3 ! RESULTS OF BRUNSWICK CONTROL ROOM RADIOLOGICAL ANALYSIS l t s 1 i - Ventilation Flow Rates (cfm) 30-Day Dose frem) Unfiltered Filtered Recirculated Whole Body Thyroid Beta Skin Inleakage Intake Flow Airborne Other Total Airborne Airborne Baseline Case 150 998 998 .002 .413 .415 .41 .039 Case 1 3000 1000 1000 .003 .413 .416 1.72 .044 I Case 2 3000 1500 500 .003 .413 .416 1.73 .044 1 j Case 3 3000 3000 500 .003 .413 .416 1.37 .045 i l, Case 4 3000 2500 1000 .003 .413 .416 1.37 .045 1
)
a i . !-{ l >I ?, J 1 f i h !) i 6 l I i i a i i d l l-10 l 4 NUS COAPOAATION 1 r- _ - _ _ _ _ _ - - , , - - , - - _ - - - - - - , - - _ _ ------------------,---,,----.a---------_---,----_-..------.-__.-.__Aa_--a_.---,----
It can be concluded therefore that the control room doses will not exceed the NRC dose criteria for any amount of inleakage ~ into the control room. These results are reasonable in light of the use of the standby gas treatment system, i.e. the standby gas treatment system exhausts the released inventory through the stack as ' an elevated plume. Consequently, the lower control room doses predicted in this evaluation.are primarily influenced i by the X/Os generated by a stack release. It should be further noted that the evaluation employed conservative methodology such as that associated with calculation of the postaccident core inventories. J 1
5.0 REFERENCES
- 1. U.S. Nuclear Regulatory Commission, 1980. NFC Action Plan Developed as a Result of the TMI-2 Accident, NUREG-0660, May 1980.
l 2. G. D. Whittier et al., ~1983, Control Room Habitability
)
, Evaluation Brunswick Steam Electric Plant (NRC TMI Action
! Plan Item III.D.3.4), NUS Corporation, NUS-3697 Revision 2,
$ February 1983.
t
- 3. 10 CFR, Part 50 Appendix A, General' Design Criteria 19, January 1, 1985.
1
- 4. R. G. Jaeger et al., 1968, Engineering Comoendium on Radiation Shielding, Volume 1, Springer-Verlag New York, Inc., 1968.
- 5. Carolina Power & Light Company, 1972, BSEP-1 & 2 FSAR, ,
Amendment 13, p. M14.1-1.
- 6. Carolina Power & Light Company, 1972, BSEP-1 & 2 FSAR, Amendment 15, p. M14.4-1.
h 1-11 NUS COADOAATION
l 1 APPENDIX A METHODS USED Id RADIOLOGICAL ANALYSIS I i t I e i I t h NUS COAAQ A ATION
1 1 1 1 APPENDIX A METHODS USED IN RADIOLOGICAL ANALYSIS f I I i
)
l NUS COAPC A ATION
i- s I , i APPENDIX A METHODS USED IN RADIOLOGICAL ANALYSIS } The control room dose calculation computer program (AXIDENT) con- )i sists of a release pathway model and a dose evaluation model. The release model computes activity inventories and releases in the containment and control room based on TID-14844 (Ref. 1) releases and prespecified flow rates, filter efficiencies, halogen non-removal factors, and meteorological data. The program computes i individual doses within the control room. A.1 RELEASE MODEL The activity release pathway model is shown in Figure A-1. Four
- activity nodes are represented
- two primary containment volumes (sprayed and unsprayed), the secondary containment volume, and the control room. The equations for nodal activities, containment re-l lease and integrated control room tactivity are derived from first l order activity balances in the following paragraphs. The defini-tions of all variables used are presented in Section A.3.
1 4 A.l.1 Primary Activity lf The primary containment activity is the sum of the activity in
- the sprayed and unsprayed regions, i
d A
- P A1 + A.,- (1) 1 1
- A,p Ag -AA y g-AAr
=
1
-A p A1 -
A # A' (2I 1 dA dt
" ~
1 2 A ~ A r 2
~
A p 2 A2* A l E A-1 I NUS CC A AC A ATICN
The simultaneous solution of Equations 2 and 3 when combined with Equation 1 gives the primary containment activity as
~
A "C (4) p 2. e 2 - C) e "It i A 10 IA 1 - *1} A 20 I 2 ~ "I} C = (5) 2 m 2 ~ "I A I
-"2)*A20 I ~ *2 }
q C = 10 1 (6) 3
*2 ~ "I i
m ,m 1 2
=
i(A 1+A 2+S+1) y 1 v 2 (7) i .
' +t (A 1 +A 2 + v1 + Ev )2
, 3 2 i -k' r i! -4(y2 x't . g x x', x;)
2 1 4 A = +A sp (5) l 1 A1+A+A r p
- i 4
A 2
= A 1
+A+A r p (9)
- i Ag = C 4
e ~m2* -3 C eT M L i 1 4 l 9. I . i 4 A-2 i NUS CCAACAATICN i
. . . =. . - . - . - _ _ _ . _
1 ' ~ A A10 ( l' *1 +V V 2 20 f C 4
=
m 1 (ll) 2 ~ *1
-m
- 1 Aio ( g 2 V
~
V ^20 C3 = (12) ' m 2 ~*1
~
i A 2
= (C2 - C 4) e "2* - (C 7 -C 3 ) eT (13)
Note that the above solution for Ap degenerates to a one-volume i problem if Asp = 0. A.1.2 Secondary Activity i The rate of change of secondary containment activity is the frac-tion of the primary activity that goes to the secondary contain-
- j ment less the removal by decay, cleanup, and leakage (or exhaust) i to the environment.
4 ! dA 8 } = fs AA - A3 As -AAr s -A s As (14) d: 1 p i
=
f, AgAp -A 4A s (15)
^ ^
A 4
" A 3 r s (16) i 1
fA C f* A) C ~ A = s 1 2 ,-m 2 * - e ~* 1* + C e 4 s A -m 2 A4 ~ "I I (17)
- 4 t
' f, At C2
+
's 1 C 1 (18) l C5 = A so A -m 2 4 A4 ~ *1 1
A-3 l! + NUS CompCA ATiON
A.l.3 Containment Activity Release Rate The containment activity release rate has two components: the secondary containment release after filtration, and the fraction of the primary containment leakage that bypasses the secondary , containment. 3 R r
= FA 3As + (1 - fs) A1Ap (19) i e
i
~
~
Al C C? 2 -m t -m t R = FA 3f sA1 .m e 2 -
^
e 1 r A 4 2 A - *1 4 - (20)
- FA3 C 5 * +
t
-m +' -m *'
(1 - f s) A l C e 2 2 -C e l 7 1 i i C e "'2 ' - C e "'1 ' + C , e _x.
! R = 4- (21) r 6 7 e
{ ~ FA f
- 3 =
j C
- I~ C I I 6 A -
s 1 2 !i . 4 "'2 . 4
'I,
~
FA, f
- C 7
=
A -m
- 1-f s A1 C 1 4 1
?
! 8
= FA 3 C 5 (24)
'- A-4
-l NUS CC A ACMTICN
I A.l.4 Integrated Release from Containment The integrated release from the containment is obtained by inte- , i grating the release rate, Equation 21, over the time period of interest. R = [R dt Jr (25) C6 C ' R = {y -m2t) _ C 7 g _,-m tj _8 ,-A4t) (26) m
- 2 *1 4 il I A.1.5 Control Room Activity i
The rate of change of activity in the control room is the dif- ,
; ! ference between the rate at which activity is drawn in from , the outside air and the rate at which it is removed by decay, i
cleanup, and leakage (or exhaust). 1;
'. q j dAc =F S ~A ^ - ^ -AA c cc 2 ec, c r r c V (27) ;- dt ec ii
!~ dA- =C R-AA (25) . 9 r i c 3 ct
= ^ * '
l A 7 A r ^^c s cc 4 C = ' (30) j 9 F,, 4 cc .VQ)c e
= C gC6 e 2-C gC7 e "*l a CgCg e t 4
; ot
-A A (31) 7 J CC CC CCo 9 -A:4 A =
96 e -m 2 t - 9 7 e
-mt
+ e c
- ! A7-m2 *7 ~*1 A7 A 4
1 A7t (32) { +C10
- i A-5 I Nt IC C.9QCN A A*'ON
. ._ _ _ _. _. ~- ._ _ _ _ .. _ . __ _
I! CC g6 CC g 7 CCg g 1 C =A - + - (33) 10 co x7
-m 2
A 7" *1 A 7 ~*4 I ; A.l.6 Integrated Activity in Control Room i The integrated activity in the control room is obtained by inte-grating Equation 32 over the time period of interest. R (34)
=[Adt 96 -m2t 97 -m)t c (A 7 - m2 }"2 1(A7 ~ *1}
Cg C g C 10
+ -
p _ x 3 (1 - 4 e_ A t) + x , (1-e -hyt) z 4 7 4 7 , J 4
- Implicit in the above derivations is the assumption of constant coefficients. In the actual transient simulation, solutiens are -
[ , broken into a sequence of discrete time intervals over which the input parameters that make up the coefficients are prespecified , constants. The input parameters consist of flow rates, X/Os, de-cay and iodine removal constants, provided as stepwise constant functions of time. I t Initial secondary containment and control room activity inventor-ies are assumed to be zero. Initial primary activity may be based on the analysis of TID-14844 '(Ref.1) using the fractional iodine l release assumptions of Regulatory Guide 1.3 (Ref. 2) or 1.4 !j (Ref. 3). The source term equation is
~
i 3 (36) A = 8.65 x 10 Pogv f,f g( 1 - e r o) (curtes) p 0 1 ' i ~ A-6 ij ,* NUS COAACAATCN l
I i - i A.2 DOSE MODEL l 1 At the end of each time interval, control room individual thy- ! roid and whole body doses are determined using the containment release rate, integrated control room activity, and input values of X/Q at the control room intake. ] js Thyroid inhalation dose in the control room is given by the { following ecuation: 1 !I i l D = D (rem) (37) T T
- 1 1 BR V
cc {Rc 1 g
, DCF i
4 i $ where j' i BR = breathing rate i ! = 3.47 x 10-4 m3 /sec (Ref. 4) I Beta dose in the control room is given by: I l D j = [ D,i i (rem) (3e1 0.23 R V s 1
, Y,1 (39)
CC 1 where l Eg = average beta energy (Mev/ dis) l (See Table A-2.) i A-7 NU9 COCDC At.* CN
i . I J Gamma dose in the control room is given by I (re=) (40) 1 D , = i[ D ,i jl = [R i [E,i,j fg pl-e# j
~
1 uj -8,J r (41) cc i ) - ! Gamma energies and f ractions are presented in Table A-1. Absorp- ! i ! tion coefficients divided by the density of air are listed in 1 j Table A-2. A.3 NOMENOLATURE 1 A Primary containment activity
- I p = Activity in sprayed volume A 1=
, A2 = Activity in unsprayed volume j' A1 = Primary containment leak rate ! Ar = Radiological decay constant (Sec-1) (See Table A-1) i Ap = Cleanup rate in primary containment i fi = Fraction of activity released to sprayed volume j f2 = Fraction of activity released to unsprayed volume
,- Vi = Sprayed volume r
V2 = Unsprayed volume 1 A3 = Secondary leak rate Spray removal rate , App
.s == Fraction of primary leakage which enters secondary l' F = Filter non-removal factor for secondary building ,
exhaust system ' i' F2 = Filter non-removal factor for control room (center) g intake system , j (X/0)e = Atmospheric dispersion to control center ' i qce = Control center intake flow !,; Vee = Control center volume i' Eri = Average gamma energy (MeV/ dis) (See Table A-21 E,31 = Average beta energy (MeV/ dis) (See Table A-2)
; Ri = Integrated release from containment (Ci)
( Ver = Control room free volume (m3) { E Energy of jth gamma of ith isotope (MeV/A) (See i i ri, j = Table A-3) , f i,3 = Fraction of jth gamma of ith isotope (T/ dis) I Energy absorption coefficient for air (m-I) (See a$ = Table A-4) { Total absorption coefficient for air (m-1) (See uj = Table A-4) i r = Radius of hemisphere with same volume as control l room (m) As = Cleanup rate in secondary containment , i
^~
)h NUS CC AAC AATION
r
- i t
i I Ac = Cleanup rate in control room 3 t
- Vee == Integrated Control center free volume (m )
control room activity (Ci-sec) j Rc i DCF{=Doseconversionfactor (rem / curie) (See Table A-2) Po = Base loaded core power (Mwt) [ Ti = Fission yield (percent) (See Table A-1) ' To = 1000 days (assumed)
- fr = Fraction of core inventory availacle for release
' = 0.25 (for iodines) (Ref. 2)
= 1.0 (for noble gases) < fi ==0.91 (for elemental iodine) (Ref. 2) 0.05 (for particulate iodine) j = 0.04 (for organic iodine)
- = 1.0 (for noble gases)
!+ 0 = Mixing flow rate between sprayed and unsprayed volumes 1 l 1i i j A.4 CALCULATION OF DOSE DUE TO DIRECT RADIATION FROM THE ; )
- BRUNSWICK REACTOR BUILCING I
l The QAD code (5) .(Ref. 5) is used to compute the integrated dose 1 i to different points within the control building from direct radiation from the reactor building after a postulated loss-of-ji ecolant accident. QAD is the generic designation for a series { cf point-kernal co=puter programs designed- for esti sting the !; effects of gamma rays and neutrons that originate in a volume-distributed source. Gamma ray dose rates, energy depositions, )i ) uncollided fluxes, and associated quantities; as well as inter-j polated moments-method neutron fluxes, energy depositions, and , dose rates, may be calculated. Surfaces, defined by quadratic equations, are used for a three-dimensional description of the physical situation. Speed, flexibility, and ease of use, as well as the ability to mock-up any direct-beam radiation problem, l contribute to the utility of the program. r i j A-9 ( NUG COPAC A ATION
The source used is based on the following equations: dA. . (1)
"~
1yA dt dA, fg ,
" A at Y 1 - 1 9- A' 4J1 where t )
A 1
=1 r +( j
( t y
. fqi Ao =Ay+ l\yt l 2
, A
= radionuclide decay constant (hr~1) r i
l = primary containment leakage rate (hr-1) N1
= 0.5 percent / day = 2.1 x 10-4 hr-1
= removal rate due to Standby Gas Treatment System h.b 1;, -
(SGTS) operation (hr-1)
= 9.0 x 10-2 hr-1 (based on an SGTS flowrate of 3000 cfm and a secondary containment volume of 2 x 10 6 f 3) ,
The solution to the above equationc is: Ay (t)y = A10 * () 1 l A-10 i NUS COAPCAAT;CN
F t r i l qsyl A
- I
\ A
- I 10 -At 10 -A t g)
A,(t)
~
= ,-- ,, e l _ e 2 A I 2~ 1[ g 2 1) where A10 is the initial activity in the primary containment i
A2(t) is the total secondary containment activity (Ci) A1(t) is the total primary containment activity (Ci) t A-11 i NUS CCAPC AAT:CN
E i A.5 REFERENCES
- 1. J. J. DiNunno et al. 1962. Calculation of Distance Factors i for Power and Test Reactor Sites. TID-14844.
- 2. U.S. Atomic Energy Commission. 1973. Regulatory Guide 1.3,
" Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors." Rev. 1, Directorate of Regulatory Standards.
- 3. U.S. Atomic Energy Commission. 1974. Regulatory Guide 1.4,
" Assumptions Used for Evaluating the Potential Radiological i Consequences of a Loss of Coolant Accident for Pressurized Water Reactors." Rev. 2, Directorate of Regulatory Standards.
- 4. International Commission on Radiological Protection. 1959.
Recort of Committee II on Permissible Dose for Internal Radiation. Pergamon Pres.
- 5. CA-3573, QAD-Point-Kernal General Purpose Shielding Codes, Oak Ridge National Laboratory.
- 6. M. E. Meek and B. F. Rider. 1968. Summarv of Fission
. Product Yields for U235, Pu 239, and Pu241 at Thermal,
, Fission Soectrum and 14 MeV Neutron Enercies. APED-5395.
- 7. C. M. Lederer et al. 1968. Table of Isotooes. 6th edition.
New York: John Wiley and Sons. - I
- 8. " Final Environmental Statement Concerning Proposed Rule Making Action: Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet th,e C:'iterion 7
'As Low as Practicable' for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents,". WASH-1258, Volume 2, Directorate of Regulatory Standards, U.S.A.E.C.,
July 1973. j1 l NUS' CC ADC A A?CN
- 9. J. H. Hubbell. 1969.' " Photon Cross Sections, Attenuation Coefficients, and Energy Absorption Coefficients from 10 kev to 100 GeV," NSRDS-NBS 29.
l l I I A-13 1 f. NUS COAPOAAT:DN
. o TABLE A-1 NUCLIDE DECAY CONSTANTS AND FISSION YIELDS (Ref. 6) Decay Constant Fission Yield l i Nuclide (sec~l) (percent)
- I131 9.97 ( 7)a 2.91 I
Il32 8.37 (-5) 4.33 Il33 9.17 ( 6) 6.69 I134 2.22 (-4) 7.8 I135 2.87 (-5) 6.2 Kr B3: 1. 0 3 (-4 ) 0.52 Kr85m 4.38 (-5) 1.3 Kr85 2.04 (-9) 0.27 KrB7 1.52 (-4) 2.5 Kr 83 6.88 (-5) 3.56 i
- Xe131m 6.79 (-7 ) 0.022 Xe l33m 3.55 (-6) 0.17 Xel33 1.52 (-6) 6.69 Xe l35m 7.40 (-4) 1.8 Xe 135 2.11 (-5) 6.3 Xel38 6.60 (-4) 5.9 aRead as 9.97 x 10-7 i
l i A-14 NUS CC APCAATKON
I
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i l TABLE A-2 AVERAGE BETA AND GAMMA ENERGIES AND IODINE INHALATION DOSE CONVERSION FACTORS Nuclide Y(MeV/ dis) (Ref. 7) $(MeV/ dis). (Ref. 7) DCF (rem / curie) (Ref. 8) I131 0.371 0.197 1.48 (+6) 1132 2.40 0.448 5.35 (+ 4 ) Il33 0.477 0.423 4.00 (+5) I l34 1.939 0.455 2.50 (+4) 7135 1.779 0.308 1.24 (+5) Kr 835 0.005 0.034
' Kr 85m 0.156 0.233 Kr 85 0.0021 0.223 Kr 87 1.375 1.050 Kr88 1.743 0.341 Xe131m 0.022 0.135 Xe l33" 0.033 0.155 Xe133 0.030 0.146 Xel35E 0.422 0.097 Xel35 0.246 0.322 Xe138 2.870 0.800 A-15 -
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a TABLE J-4 ABSORPTION COEFFICID'1S FOR AIR (Ref. 9) E u/, (a ) Ma/o (b) cm2/gm MeV
- 0.01 4.99 4.61 l l. 0.015 1.55 1.27 0.02 0.752 0.511 0.03 0.349 0.148 j 0.04 0.248 0.0669 0.05 0.208 0.0406
- 0.06 0.188 0.0305
-
- 0.08 0.167 0.0243 0.1 0.154 0.0234
, 0.15 0.136 0.0250 C.? 0.123 0.0268 i Lc1 0.107 0.0288
' - 0.4 0.0954 0.0295 0.5 0.0870 0.0297 0.6 0.3805 0.0290
) 0.8 ' . 0707 0.0289 1.0 0.v!'6 0.0280 1.5 0.051s 'J.0257 2.0 0.0445 0.0238 3.0 0.0358 0.0212 4.0 0.0308 0.0194 j aF cm Table 3.-27, NSRDS-NES 29. bFrom Table 1.-7, NSPSS-NBS 29. W e-I I. L A-18
- NUS CO ACC AATCN
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