ML20246L039
| ML20246L039 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 07/14/1989 |
| From: | HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20246K903 | List: |
| References | |
| NUDOCS 8907180330 | |
| Download: ML20246L039 (37) | |
Text
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ATTACHMENT O l
PAGE 1
Oc 2 7 6.4.3 System Operational Procedures l
l-The method of operation of the control room envelope HVAC system during normal
..l 36 '
, plant and emergency conditions is described in Section 9.4.1.
i 6.4.4 Design Evaluation Each of the operating systems which ensures control room envelope habitability l36 is discussed in detail in other sections. These systems and the sections in which they are discussed are as follows.
l38 l52 Electrical Auxiliary Building HVAC Systems 9.4.1 1
Fire Protection System 9.5.1 Communication System 9.5.2 Lighting System 9.5.3 52 offsite Power System 8.2 Onsite Power System 8.3 Radiation Monitoring System 11.5, 12.3.4 6.4.4.1 Radio 1ctical Protection. The control room envelope HVAC. system l36 has been designed to minimize the airborne radioactivity dose to the plant operators after a Design Basis Accident (DBA).
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The postulated 14CA has be'en qualitatively determined to be thw MA resulting in the highest control room operator doses. The radioactive transport model 38 is described in Section 15B. Refer to Section 15.6,5 for a discussion of the offsite environmental consequences of a postulated IhCA, using the assumptions of RG 1.4.
These assumptions have also been used for the control room dose analysis. Due to the close proximity of the charcoal filters to the control room envelope, the filter units dose contributions were considered in the con-38 trol room envelope dose analysis.
The emergency HVAC for the control room envelope is discussed in Section 9.4.1.
The system configuration is shown on Figure 9.4.1-2.
The mathematical model used to represent the system uses a single outside air intaka (Ref.
9.4.1-2) and a filtered pressurization ir, flow which mixes with part of the return air, and then the mixed air is filtered again before being supplied to M
G he air handling unit along with the remaining return air,' The air handling 36 A
unit supplies the conditioned air to the control room envelope. A summary of these parameters is presented in Table 6.4-2.
An unfiltered inleakage of 10 sefm to the control room envelope has also been assumed (Ref. 6.4-4).
[38 The atmospheric releases from the Containment purge va or to closure and from the FHB (ESF leakage) are assumed to be transport the control room l36 envelope air intake by the atmospheric (meteorological) conditions existing at the time. These conditions are estimated using the methods of Reference 6.4-4.
The atmosphere dispersion factors for each case can be found in Table 6.4-2.
B907100330 890714 PDR ADOCK 05000498 6.4-6 Amendment 52 P
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ST-HL-AE-ai4 0 0F 37 PAGE 3
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If one; train of control room HVAC is inoperable, for example due to diesel l3 ccnerator f ailure, not all of the makeup air would be filtered twice j#
b2 fore it is introduced into the control. room envelope.
In the worst
'coce,1235 cfm of the makeup air is filtered by the makeup units, but not by the~ recirculation. units, bef ore it.is introduced into the control room.
envelope.
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During the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the control room envelope is T
considered to be occupied continuously. Between 1 and 4 days after the acci.
dent, a 60 percent occupancy is assumed. Following 4 days after the accident o.c4 a 40 percent occupancy is assumed for the remainder of the accident. The brea-
'" D thing rate used for the operator is 3.47 x ION s
a /sec (Ref. 6.4-4 and RG V
1.4).
4 The inhalation thyroid dose an( zhe semi-infinite cloud gamma and beta doses are calculated using the time-integrated concentrations in each area and the occupancy factors noted in Table 6.4-2.
The semi-infinite. cloud model remains appropriate caly for the beta dose, due to the short range of beta particles.
The semi-infinite cloud gamma dose calculated is divided by a geometric factor which converts the semi-infinite gamma dose to a finite dose (Ref. 6.4-4).
This factor is given as:
gp,1173-0.338 y
where:
V - volume of region of interest, ft:
The resulting doses to control room personnel are given in Tab'le 6.4-2.
The calculated thyroid dose total is less than the design limit of 30 rem, as is the skin beta dose total.
The total whole-body gamma dose is less than the design limit of S rem.
Thus the control room envelope HVAC Systen, design meets the dose requirements of GDC 19 of 10CFR50, Appendix A.
6.4.4.2 Toxic Cas Protection. The general guidance contained in RG 1.78, has been considered in the design of the control room envelope HVAC system, as described in Section 9.4.1.
Toxic gases which are handled onsite are kept to a minimum. During normal operation small amounts of chlorine are handled within the site boundary at the Training facility. The amount of chlorine (<300 lbs) will not impact the control room envelope. A detailed evaluation of potential hazardous chemical accidents and their impact on control room habitability is provided in Section 2.2.3.
6.4.5 Testing and Inspection Systems and their components, listed in Secticn 6.4.4 above, which maintain control room eovelope habitability are subjected to documented preoperational testing and inservice surveillance to ensure continued integrity.
The tests conducted vertfy the following for both normal and emargency conditions.
1.
System integrity and leaktightness 2.
Inplace testing of emergency filter plenums to establish leaktightness of plenums and design parameters of the high efficiency particulate air and charcoal filters 3.
Proper functioning of system components and control devices i
4.
Proper electrical and control wiring 6.4-7 Amer.dment 53
c 1
STP FSAR ATTACHMENT 3 ST-HL-AE-394 o i
PAGE 4 0F 77 TABLE 6.4-2 4
CONTROL ROOM DOSE ANALYSIS I
l Assumptions l
Containment leakage assumptions 0.3% (0-24 hrs) 0.15% (1-30 days)
ESF system leakage assumptions 8280 cc/hr Pressurirstion make-up air inflow parameters:
flowrate 2,000 ft s/ min filter efficienc
- 7#.5 49% incrganic, i
l W, 5'
- organic, 9
articulate Control room envelope clean-up air (recirculation) parameters:
filtered flowrate 10,000 ft 8/ min (recirculation air) filter efficiency 95% inorganic, 95" organic, 99% particulate 2.13,060 3g 2 L ^ ^^ f tS envelope free volume envelope unfiltered inleakage 10 ft 8/ min Meteorological dispersion factors (including vind speed and Ess t%k c.a 3
direction allowances):
&qs Gse.
Containment EST =d Peete Leakage Case kde;;: "-M
~8 sec/m*
0-8 hours 1.06x10jsec/m' 1.29x10 8-24 hours 7.03x10 sec/m8 8.55x10~5 sec/m'
~
~4 1-4 days 4.45x10 sec/m' 5.42x10 s,,ef,s
~4
~8 sec/m' 4-30 days 1.91x10 sec/m' 2.32x10 Occupancy assumptions:
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, control room 100%
1-4 days, control room envelope 60%
4-30 days, control room envelope 40%
Breathing rate of operator 3.47 x 10 ' m'/see
~
h enkeup & resiecula4 ion 4))ses t1765 ek is $$$sreb 0"%
AHus Q. EffecAiw AHer e(Acte ~q (,. -as ef,., is w2 boo c{n
&Ihrul Krey k n k wy eS gi m O ove.
6.4-12 Amendment 38
'~iii UHMENT3 ST-HL-AE-1910 l
TABLE 6.4-2 (Continued)
CONTROL ROOM DOSE ANALYSIS-Results Whole-Body Skin Operator dose. 0-30 day period (rea):
Thyroid Causna Beta Containment leakage 1.5 8.7,4 M
i. H s'-h49 10'5 1
ESF leakage 1
10 Containment purging
'q.046-Gr0%
x10-5
.2x10*'
direct dose from Containment
- 0.11 direct dose from cloud of
{aS3M released fission products 51 lodine filter loading 2.21x10,3 38 Total 18.
50 1
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6.4 13 Amendment 56 l
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Ecuron.nse c&s STP FSAR Poa yhv7 eHowa sa
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TABLE 6.5-2
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INPUT PARAMETERS AND RESULTS OF SPRAY IODINE REMOVAL ANALYSIS
@h W
Power (102% of rated core power) 3876 MWt I
Containment pressure 37.5 psig 5 6 Containment temperature 307 F 38 6
3 Total containment free volume 3.56 x 10 gg Unsprayed Containment free volume 23%
Spray fall height 136 ft Net spray flow (3 pumps) 4530 gal / min I}
Spray solution pH Half flow at pH - 8.0 56 Half flow at pH - 9.0 8.8 Elemental iodine A,( )
.I Note:
(1) The limiting single failure is that of a spray additive tank discharge valve failing to open which can result in two separate pH levels simultaneously due to the three spray pumps discharging to two headers.
56 h)
The value of A given here represents its calculated value; however, offsile dose calculations utilize a more conservative value (see Table 15.6-10).
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6.5-15 Amendment 56
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{ /apca.sei h Q g';~ 5 STP FSAR Appendhc 7A h
STP Position Q$
A review of the post-accident radiation environment for both access and equip-ment qualification has been performed using the methodology and assumptions described below.
I
-Source Terms 1-The core inventory for STP was generated using a 3 region core model (300, 1
600, 900 EFPD) with a conservative core power level of 4100 MWt.
This core inventory was partitioned as follows:
45 Airborne Source:
1004 Noble cas, 504 Halogens Liquid Source:
504 Halogens, it folids Each source was diluted by the appropriate dilution volume.
In the airborne case this was the containment net free volume while in the liquid case it was
'the. total liquid volume of the primary system, accumulators, and the available volume of the refueling water storage tank.
The airborne source was assumed to be instantaneously released and distributed i
throughout the containment atmosphere.
In the liquid case the source was decayed.for a short period equal to the time required for recirculation to begin.
It was assumed to be distributed in the containment sump (no decay assumed), portions of the Residual Heat Removal, Safety Injection and Containment Spray systems.
Post-Accident Radiation Zones Using the source terms described, radiation zone maps were generated for the Reactor containme t Building, Mechanical and Electrical Auxiliary Building, Fuel Handling Bu1A'ing, and the Isolation valve Cubicle. The resulting zone cape can be found in Section 12.3 (Figure 12.3.1-19 thru 36) for the time t-53 after the accident.
1 (1) Continuous occupancy 45 The areas requiring continuous occupancy, control Room Envelope (Section 6.4) cnd the Technical Support Center, were found to have sn average dose rate less j
than 15 mR/hr for the 30 days following the accident. The thirty day inte-grated doses were determined and found to be below CDC 19 limits (Table 53 II.B.2-2).
(2)
Infrequent Access The infrequent access areas were reviewed in conjunction with the use of the l
post accident radiation zones found in Section 12.3.
Using these drawings, a g3 review was made of the routes used to reach each area and the expected dose rates at each location was analyzed. Based on this review, each of the areas were found to be accessible from the control room. The area dose rate at t
7A.II.B.2-4 Amendment 53
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E.
STP FSAA App;ndix 7A wpgg gy
-i various times, for each location, after the accident has been provided in Table II.B.2 2.
In the event that entry is. required in these areas, due con-53 sideration is given to the dose rates expected and appropriate time limits for presence in the area are imposed to encure that the doses received will not-exceed CDC 19 limits.
Radiation Qualification of Safety Related Equipment The same source terms, described above, were employed in obtaining the post 45 accident qualification doses.
(LOCA doses were found to bound the high energy line break doses).
Further discussion and the results of the analysis can be 7
found in Section 3.11.
Table II.B.2-1 summarizes these considerations.
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,.. u.GriiEENT 3 ST-HL-AE-2740 STP FSAR PAGE 9
__0F 3 7 Appendix 7A TABLE II.B.2-2 Post Accident Radiation IAvels/ Doses Continuous Occupancy Areas:
30 day Doses (Rem)
Camma Beta Thyroid Control Roon
.2,4y2 44 18.70
.146 IT.08 h
Technical Support Center A-98'4.Bo 31,4221.f4 19. 55 '2.8.40 Infrequent Access Areas:
FSAR Figure Dose Rate (R/Hr)
Reference Area Time after accident I hr 1 day 1 wk 1 month 12.3.1 36 Post-accident
.75 4.5 x 10 s 1,1 x to 6 x 10'*
sample station 12.3.1-27 Health Physics 6 x 10 s 3.6 x 10"*
9 x 10~5 4.8 x 10 s counting room 12.3.1-27 Radwaste count-3.1 x 10 s 1,3 x to s 4.6 x 10~4 2.4 x 10~5 ing room 12.3.1-28 Flant vent 4.74
.28 7.1 x 10~8 3.8 x 10 s radiation monitor 12.3.1-25 Auxiliary shut-8 x 10"*
4.8 x 10~5 1.2 x 10~5 6.4 x 10
down panel 7A.II.B.2-7 Amendment 53 l
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'~hflACHMENT3 Pnac nwsoco rox.
l ST-HL-AE-:@o PAGE 3 0F 37 STP FSAR INFowW e 7 (drawing water from the sumps) to further reduce Containment pressure.
Approximately 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> af ter initiation of the LOCA the ECCS is realigned to l45 supply water to the RCS hot legs in order to control the boric acid concentra-tion in the reactor vessel.
Description of Small Break LOCA Transient As contrasted with the large break, the blowdown phase of the small break occurs over a longer time period. Thus, for the small break LOCA there are only three characteristic stages, i.e.,
a gradual blowdown in which the decrease in water level is checked, core recovery and long-term recircula-tion.
A block diagram summarizing various protection sequences for safety actions 2 l45 required to mitigate the consequences of this event is provided in Figure 15.0-25.
g3 3, 6
15.6.5.3 Environmental Consequences. The results of analyses presented in this section demonstrate that the amounts of radioactivity released to the environment in the event of a postulated LOCA do not result in doses which exceed the limits specified in 10CFR100.
Dose contributions from three different sources are considered:
Containment l45 leakage, leakage from Engineered Safety Feature (ESP) components, and purging of the Containment prior to isolation. The parameters used for these analyses l45 are summarized in Table 15.6 10.
15.6.5.3.1 Containment 1,eakage Contribution:
Following a postulated double-ended rupture of a reactor coolant pipe with subsequent blowdown, the ECCS keeps cladding temperatures well below melting, and limits zirconium-water reactions to an insignificant level, assuring that the core remains intact and in place.
As a result of the increase in cladding temperature and rapid depressurization of the core, however, some cladding failure may occur in the hottest regions of the core.
Thus a fraction of the fission products accumulated in the pellet-cladding gap may be released to the RCS and thereby to the Containment.
15.6.5.3.1.1 Activity Release to Containment - The offsite doses result-l45 ing from a hypothetical accident releasing core activity have been analyzed.
Activity releases of these magnitudes have a considerably lower probability than that associated with gap activity releases.
For the analysis of this hypothetical case, the assumptions outlined in RG 1.4 were used to determine the initial activity release.
Thus a total of 100 percent of the core noble l45 gas inventory and 50 percent of the core iodine inventory is assumed to be immediately available for leakage from the Containment.
The total core
!33 inventory is given in Table 15.A-1.
Of the iodine activity released to the Containment, it is assumed that 95.5 l45 percent is in the elemental form, 2 percent is in the organic or methyl 51 iodine form and 2.5 percent is in particulate form.
(
)
15.6.5.3.1.2 Containment Model Parameters - The quantity of activity l
released through leakage from the Containment was calculated with a two-volume I
model of the Containment to represent sprayed and unsprayed re6 ons of the 1
Containment.
This model is discussed in Appendix 15.B.
15.6 13 Amendment 51
hlAGNNIENT'5 ST-HL-AE-Wf o STP FSAR PAGE o n OF 7 7 -
/
The containment leak rate to the atmosphere used in the analysis is the design l45 basis leak rate "!? r!.1 h indicated in the Technical Specifications.
For.
the first-24 hours following the accident, the leak rate is assumed to be 0.30 'l p percent per-day while for the remainder of the 30 day period the leak rate is assumed to be 0.15 percent per day. This Containment leakage is assumed to leak directly to the environment, s,%
The tot 31 free volume of the Containment has been calculated to be &-64 mil-l45 lion ft Part of this volume is covered by the Containment spray, while some is not.
The major portion of the unsprayed volume is within the secondary -
shieldwallbelowtheoperatingf}oor. The unsprayed volume has been calcu-lated as approximately $lo ooo84&;409 ft l45 s
The transfer rate between the sprayed and unsprayed regions is assumed to be limited to the forced convection induced by the Reactor Containment Fan Cooler.
(RCFC) units. The number of units assumed in operation and the total mixing
' flow are presented in Table 15.6 10.
This assumed minimum flowrate conserva-tively neglects the effects of natural convection, steam condensation and diffusion, although these effects are expected to enhance the mixing rate between the sprayed and unsprayed volumes. The majority of the RCFC air sup.
ply, except a small portion discharged to the dome, is discharged to the space within the secondary shield wall, where it is relieved to the balance of the Containment volume through the vent areas. The RCFC units are described more fully in Sections 6.2.2 and 6.2.5.
For fission products other than iodine, the only removc1 processes considered are radioactive decay and leakage.
lodine is assumed to be removed by radio-active decay and leakage, plateout, and also by the Containment Spray System l56 (CSS). The effectiveness of the Containment spray for the removal of the iodine in the Containment atmosphere and the model used to determine the iodine removal efficiency are discussed in Section 6.5.2.
Only the elemental A spray removal rate of W hr g.is assumed, until the airborne h2 l56 and particulate iodine forms are, assumed to be effectively removed by the
/8,8 spray.
g2.,f elemental lodine is reduced by a factor of"M-96 fter this time. the,(e g 56 elemental spray removal rate is assumed to be ze eached.gplateoutand wever.
p6rticulate removal continues until a DF of 100 The sprays are considered effectiv o y in the sprayed region of the Containment.
15.6.5.3.1.3 conta e
Leakage Doses - Doses resulting from activity leakage from the Containment have been calculated using the models presented in Appendix 15.B.
The thyroid, whole body gamma and skin beta doses are presented in Table 15.6 11 for the Exclusion Zone Boundary (EZB) distance of 1,430 meters and the outer boundary of the Low Population Zone (LPZ) at 4,800 j
meters.
d 15.6.5.3.2 ESF Leakage contribution: A potential source of fission product leakage following a LOCA is the leakage from ESF components which are located in the Fuel-Handling Building (FHB). This leakage may be postulated to occur during the recirculation phase for long-term core cooling and Con-tainment cooling by sprays. The water contained in the Containment sumps is used after the injection phase and is recirculated by the ECCS pumps and the l45 Containment spray pumps.
I 15.6-14 Amendment 56
-=
a
.c ST HL-AE-:t9yo STP FSAR PAGE 31 OF M
=15.6.5.3.2.1 Fission Product Source Term - Since most of'the radiolodine released during the LOCA would be retained by the Containment sump water, due to operation cf the CSS and the ECCS, it is conservatively assumed that. 50 percent of the core iodine inventory is introduced to the sump water to be recirculated through the external piping systems.
l45 l'
Because noble gases are assumed to be-availabIe for leakage from the Contain-
):
ment atmosphere and are not readily entrained in water, the noble gases are not assumed to be part of the source term for this contribution to the total.
LOCA dose.
l 15.6.5.3.2.2 Leakage Assumptions - The amount of water in the Contain-I >
ment sumps at the start of recirculation is the toi 1. of the RCS water and the water added due to operation of the engineered safeguards, i.e.. the ECCS and CSS. This amount has been calculated to be 512,494 gallons. This value is l45' conservatively low to maximize iodine concentration in the sump water.
The ECCS recirculation piping and components external to the Containment are designed in accordance with applicable codes and are described in Se : tion 6.3.
The CSS is described in Section 6.2.2 and 6.5.2.
45 The maximum potential recirculation loop leakage is tabulated in Table 15.6-12. ' Each recirculation subsystem includes a high-head safety injection -
(HHSI) pump, a low head safety injection (LHSI) pump, a residual heat exchanger, the Containment sump, and associated piping and valves. Thus three separate subsystems are provided for recirculation as well as for inject on, each of which is adequate for long term cooling.
Since three redundant subsystems are available during recirculation, leakage for any component in any subsystem can be terminated by shutting down the LHS1 and HHSI pump associated with that subsystem and by closing the appropriate pump suction and discharge isolation valves.
l45 Maximum potential recirculation leakages are indicated in Table 15.6 12.
The l45 leakage rate assumed for dose calculation purposes in conservatively twice the leakage rate given in Table 15.6 12.
The iodine partition factor applicable for this leakage is assumed to be 0.1.
l45 I
l 15.6.5.3.2.3 ESF Leakage Doses - The iodine activity, once released to the atmosphere of the FHB, is assumed to be quickly transported by the venti-lation system through the exhaust filters and released to the environment at ground level. The iodine filtration efficiency is assue d to be 95 percent.
l45-
[
The offsite doses due to the recirculation leakage are presented in Table 15.6-11 for the EZB of 1,430 meters for the initial two hour period and the LPZ outer boundary distance of 4,800 meters for the 30 day duration of the accident.
15.6.5.3.3 Containment Purge Contribution: In the event of a LDCA coin-cident with the containment supplementary purge system in operation, the purge is assumed to be isolated within 23 seconds following LOCA initiation. During normal power operation,3the containment supplementary purge system vents the containment at 5,000 ft / min. However, for this analysis the maximum flow 45 e due to the pressure spike inside the Containment was used (88,900 rag / min).
ft The containment purge system is described in Section 9.4.
.fw..cf 15.6 15 Amendment 56 Ihs, l.dsks a d n
- shsust-
L4 C
Mi4CEii}ENT 3 L
ST-HL-AE-1WO PAGE 3 0F M 7 l
YNSERTA*
PAGE :3.6-15
(-
. For'the first 30' minutes, the analyses' consider the potential _ imbalance of exhaust flow, resulting in flow through one train of heaters being below thn setpoint f or heater. energization (12,000 cfm through'the three filter bcnks).
During this time, the filter ef ficiency is assumed to be 90 percent f or elemental iodine and 30 percent for organic iodine for that trcin of filters.
Within 30 minutes, operator action to isolate that trcin of filters occurs.-
1 e
'"iE h The containment airborne iodine inventory av 11able for release is assur d to I
j be the flashed portion of tho total primar coolant iodine inventory ba..d on
.pg a pre existing iodine spike level of 60 ci dose equivalent I-131.
For
-.D noble gases, 100 percent of the primary coo ant inventory based on 1 percent 45-gg failed fuel is assumed to be available for release. No failed fuel is assumed since isolation occurs prior to the core reaching a temperature which could b
cause a fuel failure.
15.6.5.3.3.1 Containment Purge Doses - The offsite doses calculated due to Containment purging are presented in Table 15.6 11 for the exclusion zone boundary (EZB) of 1,430 meters and low population zone (LPZ) outer boundary distance of 4,800 meters.
g 15.6.5.4 Core and System Performance.
15.6.5.4.1 Mathematical Model:
The requirements of an acceptable ECCS Evaluation Model are presented in Appendix K of 10CFR50 (Ref.15 6-2).
Large Break LOCA Evaluation Model The snalysis of a large break LOCA transient is divided into three phases:
1) blowdown, 2) refill, and 3) reflood.
There are three distinct transients analyzed in each phase, including the thermal hydraulic transient in the RCS, the pressure and temperature transient within the Containment, and the fuel and clad temperature transient of the hottest fuel rod in the core. Based on these considerations, a system of inter-related computer codes has been devel-oped for the analysis of the lhCA.
The description of the various aspects of the LOCA analysis methodology is given in WCAP-8339 (Ref. 15.6-4).
This document describes the ma.Jor phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with the Acceptance Criteria. The SATAN VI, VREFLOOD, COCO, and LOCTA-IV codes, which are used in the LOCA analysis, are described in detail in References 15.6-5 through 15.6-8.
Modifications to these codes are specified in References 15.6 9, 15.6-10, and 15.6-11.
The BART code is described in References 15.6-11e and 15.6-11f.
51lIO These codes are used to assess the core heat transfer geometry and to determine if the core remains amenable to cooling throughout and subsequent to the blowdown, refill, and reflood phases of the LOCA. The SATAN VI computer code analyzes the thermal-hydraulic transient in the RCS during blowdewn and the VREF1h0D computer code is used to calculate this transient during the refill and reflood phases of the accident. The COCO computer code is used to calculate the Containment pressure transient during all three phases of the lhCA analysis.
Similarly, the LOCTA-IV computer code is used to compute the thermal transient of the hottest fuel rod during the three l
phases.
l l
The large-break analysis was performed with the approved December, 1981 version of the Evaluation Model (Reference 15.6-10), with BART (Reference 51 15.6-11g) which includes modifications delineated in Reference 15.6-11h.
i 15.6-16 Amendment 53 q
l I
i l
l L.
l
7
.I
' ~Aii A5sII5iiT 3-1 1
STP FSAR ST-HL-M-494 0 PAGE E OF T1 L
TABLE 15.6-10 l
PARAMETERS USED IN ANALYSIS OF l.
LOSS-OFf-COOLANT ACCIDENT OFFSITE DOSES t
Parameter Core thermal power, MWt 4,800 -he0G Containment model 2 volume (spray and unsprayed)
Activity released to containment and available for leakage' 100% core activity
- noble gases Table 15.A-1 a
50% core activity 151 iodines Table 15.A-1
(
Form of iodine activity elemental 95.5%
51 56 organic 2.0%
particulate 2.54 3
containment free volume, ft 3.ss 6
total
-}.-M x 10 unsprayed 46 x 10 45 8.20 Containment' leakage rate, t per day 0 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.30 1-30 days 0.15 Number of RCFC units operating 3 of 6 Hixing rate between g/ min prayed and 160,500 unsprayed region, ft Renoval-coefficien{s 19,3 l56 elemental thp_ -(1,y, +
organic, hr 0.0 1
^plateout,hr~[
particulate, h 6.13 4.ol 51 unsprayed
. 06 '.. '
e lema.kl sprayed - 444-034 Assumed iodine DF iz,, 3 elemental (spray stops at a DF of HHB4) 100 56 organic particulate 100 s
15.6-33 Amendment 56 i
m.m_____
__m_m__
___m_____mm._____
__mm____
. ATTACHMENT 3 ST-HL-AE-Acivo E
STP FSAR TABLE 15.6 10 (Continued)
PARAMETERS USED IN ANALYSIS OF t
LOSS-OFf-COOLANT ACCIDENT OFFSITE DOSES Parameter Spray additive delivery to Containment:
2.34 51 time after initiation of 14CA, minutes Activity assumed mixed in Containment sump water available for ESF leakage i-noble gases None iodines 50% core' activity Table 15.A-1 ESF system leakage rate assumed, cc/hr Twice that of Table 15.6-12 Amount of water in which mixing of 45' iodine occurs, gallons 512,494 Iodine partition factor for leakage 0.1 FHB filtration efficiency, percent 95 Supplementary purge rate, scfm 88,900 f (4 sac.L e4 k. has s, i Ake.
J a,.kd)
Time before isolation of purge, seconds 23 56 Meteorology' 5 percentile Table 15.B-1 Dose model Appendix 15.B 4
15-6.34 Amendment 56
4 AiiactiMENT3 STP FSAR SI-HL-AE-Wo PAGE_.%
Op 27 TABLE 15.6-11 DOSE RESULTING FROM 1ARGE BREAK LOSS-OF-COOLANT ACCIDENT Parameter Containment Leakaje Doses Exclusion Zone Boundary
- 0 2 hr 2
a thyroid, rems
- 1. 2 ', ', a 10 f,4g x,o whole body gama, rems 1+
2,25 skin beta, rems 14 f/3 Low Population Zone
'.i a_10,3 g,9 yfo'8 g
skin beta, rems 4Tx 10' ESP Leakage Doses
___~___D Exclusion Zone Boundary *'O+2 hr 3
10'b 2.SIX/O thyroid, rems 2.10 a
+
- 8. oY. lo* #1 0.01 10'4 whole-body gammagrems 1
e i
1.04 a 10' p,pg xfo -V skin beta, rems Low pulation Zone
-30 days hyroid ems
^ 01 a 10' 3 69 XIO ~#
10'[ /,36x /o*4 10 3,9 / k'/o ~#
whole-body gamma, rems 3.73 a skin beta, rems t
1.3 a I
containment Purging Doses Exclusion Zone Boundary 0-2 hr I
thyroid, rems 1-7-09
/ '7,'/ 4 C.0 x 10,'3
-3 whole-body gamma 3 rems
- 3. 2 a 10 9,qq vfo kin beta, rems ?
4,g6 v/o 3 Lo ulation Zone 0-30 days hyroid, rems t Q.14 51 45 whole-body gam a, rems
.16 x 10,3 4
skin beta, rems 8.4 x 10 Total Doses l
Exclusion Zone Boundary 0 2 hr 2
thyroid, rems 1.43 x 10
/. 68 8
/, / 9 51 Low pulation Zone 0 30 days g
f thyroid, rems 0.12 a 10 4,,7 VX/O whole-body gam a, rems 0, g,8f skin beta, rems 0.43 l
- Exclusion Zone Boundary is at 1,430 m.
Outer boundary of Low Population Zone is at 4,800 m.
15.6 35 Amendment 56
ATTACHMENTS STP FSAR ST-HL-AE-AYf 0 PAGE T7 0F 31 i
taru n uaeay 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT A number of events have been postulated which could result in a radioactive release from a subsystem or component. The events which have been postulated are:
1.
Waste gas system failure 2.
Postulated radioactive releases due to liquid-containing tank failure 43 (release to atmosphere) 3.
Postulated radioactive releases due to liquid-containing tank f ailure (ground release) h3 4
Design basis fuel handling accidents 5.
Spent fuel cask drop accident The above events are considered to be American Nuclear Society (ANS) Condition III events, with the exception of the fuel handling accidents, which are con-sidered to be Condition IV events.
15.7.1 Waste Gas System Failure 15.7.1.1 Identification of Causes and Accident Description. The Gaseous Waste Processing System (CWPS) is designed to remove fission product gases from the reactor coolant and other miscellaneous sources and process these gases before they are released to the environment.
The GWPS processes these gases through a guard bed, two charcoal delay tanks, and a high-efficiency particulate air filter before release, providing delay time for noble gas activity and ample charcoal for iodine removal.
The Reactor Coolant Vacuum Degassing System (RCVDS) is designed to remove fission product gases from the Reactor Coolant System (RCS) free space prior to reactor head removal for refueling operations.
The RCVDS stores these gases in decay tanks, providing sufficient delay time for decay of noble gas 43 and iodine activity before release to the environment.
Gaseous releases from the GWPS or the RCVDS may occur or be postulated to occur as a result of leaks in piping, leaks in vessels and other equipment, and failure of vessels or other equipment. The most limiting of these is the l
rupture of a GWPS charcoal adsorber tank, providing a large break area for l43 release of activity.
The CVPS guard bed and the two charcoal adsorber tanks are designed te seismic l 43 Category I requirements; the remainder of the system is nonseismic. The Mechanical Auxiliary Building (MAB), in which the equipment is located, is, l
I however, a seismic Category I structure.
The GWPS system is classified as non-nuclear safety.
The design parameters and description of the GWPS are presented in Section 11.3.
Equipment and tanks are designed for significantly higher temperatures and pressures than are expected during operation.
Because of the conservative design of the GWPS components, an uncontrolled and unexpected rupture of a 15.7-1 Amendment 43 t
m
~~
M fo/ M nod M y ML, c
tank is considered improbable. A waste gas system failure is classified as an 3
ANS Condition III event, infrequent fault.
WPk N
A block diagram summarizing various protection sequences for safety actions 2
gg required to mitigate the consequences of this event is provided in Figure Q211.
15.0-26.
6 d4 15.7.1.2 Analysis Assumptions. Both the GWPS and the RCVDS vere analyzed. The GWPS was determined to be the Ifmiting case. The accident conservatively assumes that a full degassing of the primary system has just occurred (i.e., the beds contain one RCS volume of iodine in addition to the everage level) and that 1 percent of the iodines in the charcoal and 100
' 43 percent of the noble gases are released.
In addition, degassing of the volume control tank is assumed to continue releasing gases through the break for 30 minutes. The parameters used for the analysis are summarized in Table 15.7-1.
The postulated tank rupture would release activity to the atmosphere of the MAB.
It is conservatively assumed for the purposes of this analysis that the entire activity released by the CWPS due to the accident is released to the l43 outside atmosphere and the environment over a 2-hour period. The meteorology-cal parameters and the dose model used are given in Appendix 15.B.
15.7.1.3 Radiological Consequences.
The doses calculated due to this postulated accident are presented in Table 15.7-1 for the exclusion zone beundary of 1,430 meters and the outer boundary of the low population zone of 4,800 meters. All doses are well within the limits specified in 10CFR100.
15.7.2 Postulated Radioactive Releases due to Liquid-Containing Tank Fail-ure (Release to Atmosphere) 43 15.7.2.1 Identification of Causes and Accident Description.
Radioactive liquid releases may occur or be postulated to occur as a result of leaks in piping, leaks in tanks and other equipment, and failure of tanks and other equipment. The most limiting of these is found to be rupture of the Recycle Holdup Tank (RHT) resulting in the release of the liquid contents to the floor l43 of the cubicle.
The RHT has a non-nuclear safety design classification, and is designed as nonseismic; however it is located in a seismic Category I structure.
43 2
A block diagram summarizing various protection sequences for safety actions 92II' required to mitigate the consequences of this event is provided in Figure 6
15.0-27.
15.7.2.2 Analysis Assumptions. For the purpose of evaluating the off-site consequences of the rupture of a storage tank all tanks in the Liquid Waste Processing System (LWPS) and the RHT were considered. The tank 43 containing the highest iodine inventory was found to be the RHT. Thus, the does consequences of a failure of one RHT is analyzed. The parameters used in the analysis are summarized in Table 15.7-2.
A tank rupture would release gaseous activity to the atmosphere in the MAB.
It is conservatively assumed that the entire activity is released to the envi-ronment in a 2-hour period following the tank rupture. The fodine activity 15.7-2 Amendment 43
,imcHMEAT3
, P4c r n u,3,, g,4,
~
4,ST-HL AE-W 97
( 7 % d m asif STP FSAR PAGE M_ 0F released to the atmosphere and the meteorological parameters used are presented in Table 15.7-2 for the analysis.
15.7.2.3 Radiological consequences. The offsite doses calculated to result from the ruptura of the RHT are presented in Table 15.7-2.
The doses l43 are seen to be a small fraction of the 10CFR100 values.
15.7.3 Postulated Radioactive Releases Due to Liquid-Containing Tank Tail-ure (Cround Release) l43 15.7.3.1 Identification of Causes and Accident Description. Radioactive liquid releases may occur or be postulated to occur as a result of leaks in 43 piping, leaks in tanks and other equipment, and failure of tanks and other equipment.
An analysis was performed to determine the worst possible tank failure based on contained activity and volume. As a result of this review, the worst activity for the radionuclides considered was the evaporator concentrates tank (ECT) for radionuclides other than tritium and the RHT for tritium.
It should be noted that both of these tanks are located in a seismic Category I struc-ture.
15.7.3.2 Analysis Assumptions.
For the purpose of this analysis, all tanks containing radioactivity were reviewed taking into consideration their specific activity as well as the tank volume. The worst total activity available for release was found to be the evaporator concentrates tank for all 53 nuclides except tritium. Due to its size the RHT contains the most activity of tritium.
Thus the Cs-137, SR-90 and 1-129 contents of the concentrates tank were assumed to be released to the groundwater. A coincidental release of just the RHT tritium contents was also assumed to provide the maximum offsite nuclide concentrations. The assumptions used in determining the activities found in Table 15.7-3 with the activities provided in Table 15.7-4. gan be The radionuclides considered were obtained by comparing the half-life verses the transit time to the Colorado River (~90 years). The resulting concentra-tions and methods of dispersion can be found in Section 2.4.13.
15.7.3.3 Radiological Consequences. The radiological consequences of this accident are presented in Section 2.4.13.
15.7.4 Design Basis Fuel-Handling Accidents 15.7.4.1 Identification of Causes and Accident Description. The design basis fuel handling accident is defined as the dropping of a spent fuel asses-bly during fuel handling, resulting in the rupture of the cladding of the fuel rods in the assembly despite many administrative controls and physical limita-tions imposed on fuel-handling operations. All refueling operations are con-ducted in accordance with prescribed procedures under direct surveillance of a supervisor.
15.7 3 Amendment 53
ATfACHMENT 3
]
ST-HL-AE4Pfo STP FSAR Epire n C o w;c o PAGE 3 0 0F 37 During refueling operations, the Normal Containment Purge Subsystem is oper-l2 ating; this system is described in faction 9.4.5.
Should a fuel handling accident occur in the Containment, the RCB Purge Isolation monitors are capa-32 l
ble of identifying that the activity release has occurred and initiating Con-Q450.
tainment isolation. The fu'nction, instrument type, setpoints, safety class, OiN and other pertinent information on the RCB Purge Isolation monitors are given 32 in Section 11.5.
Isolation of the Containment is deteribed in Section 6.2.4
- Q450, which discusses the valves, mode of operation, closure time, and other infor-Oly mation.
The Fuel-Handling Building (FHB) Ventilation System is described in Section 9.4.2.
Should a fuel handling accident occur in the FHB, the spent fuel pool ventilation monitors are capable of identifying that the activity release has taken place, diverting the building exhaust flow through the carbon filter units, and starting the booster fans. The spent fuel pool ventilation meni-tors are discussed in Section 11.5.
The design basis fuel handling accident is classified as an ANS Condition IV event, limiting fault.
A block diagram summarizing various protection sequences for safety actions 2
required to mitigate the consequences of this event is provided in Figures Q211.
15.0-28 and 15.0-29.
6 15.7.4.2 Analysis Assumptions. The assumptions postulated in the calcu-lation of the radiological consequences of a fuel handling accident in the FHB or the RCB are consistent with the assumptions of Regulatory Guide (RG) 1.25.
1.
Fission Product Inventories 32 l Q450'.
The discharged fuel assembly with the peak inventory is the assembly assumed 01N to be dropped. The assembly inventory is determined assuming maximum full-power operation at the end of core life immediately preceding shutdown.
A decay period of 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> is applied.
43 The gap model discussed in RG 1.25 is used to determine the fuel-clad gap activities. Thus, 10 percent of the total assembly iodines and noble gases, except for 30 percent of the Mr-85, is assumed to be in the fuel-clad gap, of The total assembly and fuel-clad gap activities at the time and reactor shut-down and the assumed time of the accident are given in Table
.7-7.
32 2.
Analysis of Consequences Q450.
O1N An analysis of a postulated fuel handling accident in both the FHB and the RCB is performed. The parameters used for the analysis are listed in Table 15.7-9.
f The assumptions for the conservative RG 1.25 evaluation are:
32
{Q450.
a.
The accident occurs 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> after plant shutdown. Radioactive decay of the fission product inventory during this interval is taken into account.
01N k
15.7-4 Amendment 53 l
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' bc Mowsen ytd.
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ST-HL-AE -W O STP FSAR PAGE 31 0F y
- ruMri./ gu* -
-b.
All the rods in one fuel assembly rupture, plus an additional 50 fuel rods assumed to be damaged by the dropped fuel assembly.
43 e
The asi.embly damaged is the highest powered assembly.in the core.
~
c.
'Ihe values for the individual fission product inventories. in the damaged assembly.were calculated based on the total core activities 32 with a radial peaking factor of 1.65.
5(
01N' d.
The minimum water' depth between the top of the damaged fuel rods and the' spent fuel pool surface or the refueling water surface is 23.ft.
15.7 4a Amendment 53
h"
m ant.;haE 6 3 L
STP FSAR ST-HL-AEWfo PAGE 32 0F Y/
l e.
All of the gap activity in the damaged rods fs released to the refu-cling water or the spent fuel pool and is assumed to consist of 10 percent of the noble gases other than Kr-85, 30 percent of the Kr-85, and 10 percent of the total radioactive iodine in the rods at the time of the accident.
32 l
f.
The noble gases released to the spent fuel pool or refueling water
- Q450, are released at ground level to the environment.
Credit is assumed INj42 for isolation of the Containment.
g.
The fodine gap inventory is composed of 99.75 percent inorganic species and 0.25 percent organic species.
tsse.
h.
The spent fuel pool and refueling vater DFs for iodine,$a taken as 32 100 in accordance with RG 1.25.
Q450.
OlN 1.
All iodine escaping from the spent fuel pool is exhausted over a 2-hour time period at ground level to the environment.. The iodines are exhausted directly to the environment until the FHB isolation dampers close, diverting the exhaust through the charcoal filters of 43 l
the FHB Exhaust Air Subsystem. The FHB Exhaust Air Subsystem is i
described in Section 9.4.2.
Redersaa ytmve'dsd in tk FMB shaust hHer kks le ce.,%I rs44.k ldlly.
- j. (Thecharcoalfilterefficiencyisassumedtobe90percentforInor-ganic iodine and 70 percent for organic iodine, according to RG j@ gg 1.25 sJkss e</ahes l<ss ;dh is c~delled im4>J 10t**ce,0.
q
'A' h
32.
i k.
The iodines escaping from the refueling water pool in the RCB are Q4}0.
I exhausted until the containment is automatically isolated. A ground
- ggg4, level release with no filtration is assumed.
1.
The O-to 2-hour accident dispersion factors given in Table 15.B-1 32 are apolicable.
Q450.
OlN 15.7.4.3 Radiological Consequences. The thyroid and whole body doses at the exclusion zone boundary and the low population zone for the design basis 32 fuel handling accidents are presented in Table 15.7-10, for accidents occur-lQ450.
ring in the FHB and the RCB.
OlN 15.7.5 Spent Fuel Cask Drop Accident In accordance with 10CFR71, the spent fuel shipping cask is designed to sus-
}43 tain a free-fall in air of 30 ft onto an unyielding surface followed by a specified puncture, fire, and immersion in water with the release of no more than a specified small quantity of radioactivity. The design of the spent fuel handling equipment limits the postulated fall of a spent fuel shipping cask to less than 30 ft, as described in Section 9.1.
43 Since spent fuel casks are designed to withstand such Icadings, the radiologi-cal consequences of this accident are not evaluated.
1 15.7-5 Amendment 43 j
- _ = _ - _ _ _ _ _ _ _ _ - - - - _ _ - _ _.
._ _ _ A
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.*e
-1 ATTACHMENT 3 ST-HL-AE-M4 0 g PAGE ~n OF'37 INSERT
- A' PAGE 15.7-5~
. It is alao -assumed that a flow imbalance. occurs in. the FHB HVAC Exhaust Cyotem. 'such that-' flow in one set of-filters ' is approximately = 12.OOO cia.
!Just below the. flow rsquired for heater operation.
Thus, that fiIter set operates at an organic iodine ef ficiency of 30 percent.
This heatsr
' f ci t ure-to-actuate condition is alarmed to the operators c.redit for-operator action to isolate that filter train is taken at 30 minutes.
An undetected failure of;one additional heater is also assumed, with renulting organic iodine filter ef ficiency of 30 percent in one of the' Lthree operating filter banks.
4 4
..h' JS STP FSAR ATTACHMENT 3 ST-HL-AE-M V O PAGE. 3'1 0F-37 TABLE 15.7-1 CASEOUS WASTE PROCESSING SYSTEM - FAILURE ANALYSES Paramet'er
' Power level, MWt 4,100 Fuel defects 1.06 CWPS total activity at time of' accident See Section 11,3 RCS Activity See Section 12,2 Activity released from charcoal:
noble gases
'100%-
iodine lg Activity released.from remainder of GWPS components ~
1004 Meteorology.
5 percentile 43 Table 15,B-1 Dose model Appendix 15.B Results
-Exclusion zone boundary (1,430 m):
Thyroid dose, rem B.S~ -6 d Whole body gamma dose, rem I.T 4,47 Skin beta dose, rem z s 4,1.6 Low population zone (4,800 m):
Thyroid dose, rem
ll l^
15.7-6 nmendment 53
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a
, 7:,(Aj..
fjfTACHMENT3 1
=
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+!
.PAGE 36 0F - 37 STP FSAR-g6 TABLE 15.7-9 L. i
. PARAMETERS USED FOR FUEL HANDLING ACCIDENTS Parameters Power level, MWt 4,100:
l53 -
Time between plant shutdowr. and accident : hr 42-RCB normal purge isolation valve closure time,:sec' 60' l43L Radial peaking factor.
1.65.
Activity!in assembly gap 'at time of accident Table 15.7-7~
RCB dilution voleme 3
l43' (4 of containment. free volume)
Damage"to fuel assembly, rods ruptured 314
~ l43 -
Form.of iodine activity released to pool' solution:
-inorganic iodine 99.75%
32:
organic iodine 0.25%
Q450..
OlN-
' Minimum-water depth between top of damage rods'and pool surface, ft 23 Decontamination' factor in pool water:
inorganic iodine 133 43 organic iodine 1
noble gases 1
Filter efficiencies. of MiB Exhaust Air Subsystem:
elemental iodine 90%
. organic iodine 704 FHB Exhaust isolation damper closure time, sec-20 FHB Exhaust isolation damper leakage, cfm 100 43 DiB dilution volume, ft8 50,000 I
15.7-14 Amendment 53 1
G.
!*?"7*
'i
~ ATTACHMENT 3 TABLE 15.B-:.
ST-HL-AE-240 -
1 PAGE ' V7 ~ OF 31 i
~
' DISPERSION FACTORi
.I g-Minimum'
- g. p%g -
Evaluatton.
Distance Time
- 1 Point from Plant-Period X/Q (sec/m )
3
~4 EZB*
1,430 m 0-2 hours 1.3 X 10
-5
-LPZ**
4,800 m 0-2 hours 3.8 X 10
-5 0
2-8 hours #
1.6 X 10
-5 8-hours
.1.1 X-10 2.4 9G
.s M-St hours 4.3 X 10 96 7t#
-6 M-694 hours 1.2 X 10
- Minimum' exclusion zone boundary is 1,430 m.
'**0 uter boundary of low population zone is 4,800 m from plant.
MF5 L 0CA wl s1s usas U.is X/Q hha 50t L PE 0~O kr N ' P*'ti"S' y
15.B-9 Amendment 43
[... 4.. 4 p
ATTACHMENT 3' i
L ST-HL-AE "LHO l
STP FSAR PAGE 36 0F 37' l'i TABLE 15.7-10 DOSES RESULTINC FROM FUEL HANDLING ACCIDENTS 1
l Accident Occurring Inside Containment h
Exclusion zone boundary (1,430 m):
3 r
Thyroid, rem 2.73 x 10*3 Whole-body gamma, rem 1.06 x 10'y Skin beta, rem 1.33 x'10' 58 Low population zone (4,800' m):
Thyroid,' rem 7.99 Whole-body gamma, rem 3.1.x 10 22 Skin beta, rem 3.9 x 10 32 Accident Occurring In FHB Q450.
43 OlN Exclusion zone boundary y
Thyroid, rem 7.84 4-46 x' 10, y Whole-body gamma, rem 2.73 e%6 x-10 Skin beta, rem 3.ss 2. 5 x 10'3 Low population zone Thyroid, rem F.6 Whole-body gamma, rem 20,%+
x 10'2 Skin beta, rem 1.os 1--43 x 10' 1
l 15.7-15 Amendment 58 1
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