TXX-3443, Forwards Design Review of Radiation & Shielding for post-accident Operations
| ML20033D113 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 12/03/1981 |
| From: | Schmidt H TEXAS UTILITIES SERVICES, INC. |
| To: | Burwell S Office of Nuclear Reactor Regulation |
| References | |
| TASK-***, TASK-TM TXX-3443, NUDOCS 8112070246 | |
| Download: ML20033D113 (36) | |
Text
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TEXAS UTILITIES SERVICES INC.
Log # TXX-3443 WM HRY AN TOWEH. DAM.AH, TEX AM WO:
Fi1e # 10010 December 3, 1981 s (# k
($[M h' M~
Mr. S. B. Burwell Licersing Project Manager 9
- { ly U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation 4-DEC4 1981* n
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Washington, D.C.
20555
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SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION DOCKET NOS. 50-445 AND 50-446
/
FSAR SECTION II.B.2; PLANT SHIELDING ro
Dear Mr. Burwell:
Please find attached a design review of CPSES radiation and shielding for post-accident operations. This study has been performed in accordance with the guidelir.es of NUREG-0737.
If you have any questions, please call.
Sincerely, H. C. Schmidt RWH:grr Attachment 8112070246 811203"1 PDR ADGCK 05000445 S
A pp d
CPSES/FSAR
~
RESPONSE TO NRC ACTION PLAN f
These vents will be operable from the Control Room and positive indication of valve position will also be provided in the Control Room.
Procedures will be developed to address the use of the Reactor Coolant Systen vents. The procedures will address infonnation available to the operator and instructions for initiating or tenninating vent usage.
This will be implenented upon staff approval.
Details of the design and supporting information required by NUREG-0737 will oe provided as soon as the design is completed.
II.B.2 PLANT SHIELDING TO PROVaDE ACCESS TO VITAL AREAS AND PROTECT SAFETY EQUIR4ENT FOR POSTACCIDENT OPERATION Action Plan Requirements:
6 "With the assunption of a postaccident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e.,
the equivalent of 50% of the core radioiodine 100% of the core noble gas inventory, and 1% of the core solids are contained in the primary coolant), each licensee shall perform a radiation and shielding-design review of the spaces around systens that may, as a result of an accident, contain highly radioactive materials. The design review should identify the location of vital areas and equipnent, such as the control room, raawaste control stations, emergency power supplies, motor control centers, and instrunent areas, in which personnel occupancy may be unduly limited or safety equipnent may be unduly degraded by the radiation fields during postaccident operations of these systems."
"Each licensee shall provide for adequate access to vital areas and protection of safety equipnent by design changes, increased pennanent or tonporary shielding, or postaccident procedural controls. The design review shall detennine which types of corrective actions are needed for vital areas throughout the facility."
II.B-3
CPSES/FSAR RESPONSE TO NRC ACTION PLAN gSESResponse A design review of CPSES radiation and shielding for post-accident operations has been performed in accordance with the guidelines of 29 NUREG-0737. The review considered the potential radiation exposure to j
operators in vital areas and to Class 1E equipnent.
The following section describes the assumptions and methodology employed in the review and a sumary of the results.
The results of the review of radiation qualification of Class 1E ecuipment will be included in FSAR Section 3.11.
J II.B.2.1 purce Tems i
j The source tenns used in this evaluation of shielding are based on the l
postulated post-accident release of radioactivity equivalent to that described in #egulatdry Guide 1.4 and TID-14844.
i The reactor core inventory of radioisotopes is derived from the following assunptions:
i 1.
Core power level of 3565 MWt.
2.
Three-region equilibriun cycle core at end of life (E0L). The f
three regions have operated at a specific power of 40.03 MW/MTU for
)
300. 600 and 900 ' FPD, respectively.
t 1
The_ following subsections describe the three specific source tenns i
employed in the evaluation.
II.B.2.1.1 Reactor Coolant The followini fractions'of core radioisotope inventory are diluted in the reactor coolant volune of 87,700 gallons.-
II.B-4.
x
CPSES/FSAR RESPONSE TO NRC ACTION PLAN 1.
100% Noble gases 2.
50% Halogens 3.
1% Remaining fissinn product inventory (solids)
' Table II.B.2-1 presents the reactor coolant source strengths as a 29 function of time.
II.B.2.1.2 Containment Atmoshpere The following fractions of core radioisotope inventory are diluted in the contaiment free volme of 2.985 x 106 ft.
3 2
1.
100% Noble gases 2.
50% Halogens 3.
1% Remaining fission product inventory (solids)
Table II.B.2-2 present P.c contaiment atmoshpere source strengths as a function of time.
I I. B. 2.1. 3 Contaiment Sep The following fractions of core radioisotope inventory are diluted in the contaiment sump water volme of 587,800 gallons.
1.
50% Halogens 2.
1% Solids Table II.B.2-3 presents the t antainment sump source strengths as a function of time.
II.B.2.2 System Review Af ter establishing the above source terms, the systems whici; may contain highly radioactive materials in a post-accident situation have II.B-5
a CPSES/FSAR RESPONSE TO NRC ACTION PLAN 4
been evaluated. The specific systems reviewed are described in the following subsections.
29 II.B.2.2.1 Containnent Direct radiation penetrating through the containnent walls and radiation streaming through penetrations was based on the source ter.n discussed in the above section II.B.2.1.2.
In addition, containaent leakage at the design leak rate (0.1 v/o per day) of the source term specified in Regulatory Guide 1.4 (see FSAR Section 15.6.5.4) is assuned. This leakage results in a radioactive cloud which contributes to the direct radiation dose in plant buildings.
I I. B. 2. 2.2 Dnergency Core Cooling System Plant shielding has been evaluated considering the Emergency Core Cooling Systen (ECCS) to be operating in the recirculation mode of ECCS operation the residual heat removal punps take suction from the containnent sunps and the safety irdection punps and centrifugal charging punps take suction from the residual heat renoval punp discharge.
Prior to the start of recirculation, the ECCS will contain water from the refueling water storage tank. Although recirculation operation is not initiated until 15 minutes to several hours afi.er the start of an accident. the shielding review has assuned that at the beginning of the accident the ECCS contains the source tenns described in Section II.B.?.1.3.
II.B.2.2.3 Residual Heat Removal System -
Plant shielding has been evaluated considering the Residual Heat Renoval Systea (RHS) contains the source tenns specified in Section I1.B.2.1.1.
II.B-6
CPSES/FSAR RESPONSE TO NRC ACTION PLAN II.B.2.2.4 Containment Spray System Plant shielding has been evaluated considering the Containnent Spray System (CSS) to be operating in the recirculation mode. Prior to the start of recirculation, the CSS will contain non-radioactive water from 29 the refueling water storage tank. Although CSS recirculation operation is not initiated until 15 minutes to several hours after an accident (if required), the shielding evaluation has assuned that at the beginning of the accident the CSS contains the source tenns described in Section II.B.2.1.3.
II.B.2.2.5 Chenical And Volume Control System The reactor coolant Chenical and Volune Control System (CVCS) is not expected to become highly radioactive in a post-accident situation because:
1.
The system is automatically isolated.
2.
The letdown system is not required for accident mitigation.
3.
Post-accident venting (degassing) capability of the reactor cooling system is provided by renotely operated reactor coolant high point vent valves (see FSAR Section II.B.1).
Nevertheless. plant shielding has been reviewed assuning the CVCS letdown system, to a recycle holdup tank and the volune control tank (VCT). contains the post-accident source described in Section II.B.2.1.1.
II.B.2.2.6 Gas Waste Processing System The Gas Waste Processing System (GWPS) is not expected to contain highly radioactive material generated following an accident since:
II.B-7
a CPSES/FSAR
~
RESPONSE TO NRC ACTION PLAN 1.
The GWPS is not required to mitigate an accident.
2.
The GWPS is automatically isolated from components containing the radioactivity generated following an accident or these components 29 are isolated from the high'y radioactive material generated following an accident.
a f
3.
Post-accident venting (degassing) capability of the reactor coolant system is provided by renotely operated reactor coolant high point vent valves (see Section II.B.1).
However, plant shielding has been reviewed assuning the GWPS contains post-accident radioactive gases from the components described in the above Section II.B.2.2.5.
II.B.2.2.7 Post-Accident Sampling System As a result of the shielding review. the existing Primary Sanple System j
described in FSAR Section 9.3.2 was found to be incapable of obtaining and analyzing highly radioactive post accident sanples without excessive personnel exposures. Therefore a new Post-Accident Sampling System (PASS) will be installed at CPSES as described in Section II.B.3.
I Plant shielding has been evaluated considering PASS (.orponents and sample lines contain either liquid sources as described in Section II.B.2.1.1 or the containnent atmosphere source described in Section II.B.2.1.2 as appropriate.
II.B.2.3 Shielding Methods Evaluation of plant shielding following an accident includes direct-radiation form the containnent, radiation enanating from the cloud -
resulting from containment leakage, and radiation from piping 'and II.B-8 L
CPSES/FSAR
~
RESPONSE TO NRC ACTION PLAN components of systems discussed in Section II.B.2.2.
Direct, scatter, and radiation streaming through penetrations was considered. FSAR Section 12.3.2 2nd response to Question 331.22 discuss the analytical methods used in the shielding analysis.
29 II.B.2.4 Design Review The plant shielding design review has identified the radiation exposure to vital areas requiring personnel access following an accident. The plant modifications resulting from this review are described in the following section.
Post-accident radiation zone drawings have been developed to aid in the plant shielding design review and in the developnent of energency procedures. These drawings are Figures II.B.2-1 through II.B.2-24.
These post-accident radiation zone drawings reflect the radiation levels in the Unit 1 and common structures arising from an accident postulated in Unit 1.
Due to synnetry, an accident in Unit 2 would result in the same radiation levels in Unit 2 and common structures.
Radiation levels arising from nonna11y radioactive sources such as the spent fuel pools, gas decay tanks,- and demineralizers are also included in the post-accident radiation zone drawings.
These areas are identified by notes to the drawings.
The Control Room Complex and the Technical Support Center (TSC) are vital areas requiring full-time occupancy during the course of an accident. The integrated dose for these ar ias is less than 5 rem whole-body, or equivalent, for the duration of the accident in accordance with GDC 19. As seen from Figures II.B.2-17 and II.B.2-18 the dose rate in these areas is less than 15 mrem /hr.
As shown by Figures II.B.2-1 through II.B.2-24, limited access _ is provided for the hot laboratory, counting room, post-ecciderit sample panel, motor control _ centers, ESF switchgear, energency power, supplies II.B-9 L
... ~
CPSES/FSAR RESPONSE TO NRC ACTION PLAN (diesel-generators and batteries), and instrunent panels. Access to these areas may be required on an irregular basis to perfonn surveillance and inspection and post-accident radiochenical analyses.
29 To meet the requirenents for post-accident sanpling, a new systen, described in Section II.B.3, is being added. This new Post-Accident Sanpling Systen (PASS) will incorporate additional shielding and/or renote operation capabilities to ensure that the dose criteria delineated in Section II.B.3 are met.
In addition, a shielded hot cell will be added to the plant hot laboratory. Post-accident sample handling, preparation and chenical analyses will be perfomed in this hot cell.
II.B.2.5 Radiation Qualification of Class 1E Equipment The post-accident radiation exposure to Class 1E equipnent located outside and inside containnent was reviewed during the shielding design rev iew. The source term for equipnent located outside containnent is described in Section II.B.2.1.1.
For Class 1E equipnent inside contaiment, the larger of either Section II.B.2.1.1 or Section II.B.2.1.2 source tenns are utilized. For equipnent located inside containnent that is potentially sensitive to Beta-radiation, the Beta-dose guidelines of NUREG-0588 are used.
The specific post-accident radiation doses to Class 1E equipnent will be included in FSAR Section 3.11.
II.B.3 POST-ACCIDENT SAMPLING Action Plan Requirements:
"A design and operational review of the reactor coolant and containnent atmosphere sanpling line systens shall be perfonned to detennine the capability of personnel to promptly obtain (less than I hour) a sanple II.B-10 I
.. = -
8 i
Table ' II.B.2-1 i
REACTOR COOLANT POST ACCIDENT SOURCES GAPNA ENERGY SOURCE STRENGTH AT TIME AFTER RELEASE (Mev/cc/sec)
Mev/ Gamma 0.0 Hour 0.5 Hour 1 Hour 2 Hour 0.8 9.52 x 101 0 5.54 x 101 0 4.49 x 101 0 3.22 x 1010 1.3 3.68 x 101 0 1.76 x 101 0 1.45 x 101 0 9.97 x 109 1.7 3.83 x 101 0 1.99 x 101 0 1.61 x 101 0 1.08 x 101 0 2.5 3.10 x 101 0 1.52 x 101 0 1.21 x 101 0 8.40 x 109 4.0 8.07 x 109 8.25 x los 5.96 x 108 3.22 x 108 5.0 2.30 x 109 1.74 x 107 1.72 x 107 1.39 x 107 6.4 7.29 x 10s 9.22 x 10 0
0 4
4 Mev/Gama 8 Hour 1 Day 1 Week 1 Month 0.8 1.49 x 1010 7.89 x 109 2.13 x 109 5.51 x 108 1.3 4.10 x 109 9.52 x 108 1.08 x 108 3.04 x 107 1.7 3.28 x 109 6.15 x 108 1.78 x los 5.18 x 107 2.5' 1.79 x 109 9.07 x 107 1.71 x 107 5.33 x 106 1.01.x 10"s 1.19 x 10s 3.40 x 104 4.0 4.73 x 1078 5.0 3.10 x 10 5.91 x 10 0
0 6.4 0
0 0
0
.4
. + -
g y.y*g g
y
i 4
Table II.B.2-2 CONTAINMENT ATMOSPHERE POST ACCIDENT SOURCES GA M ENERGY SOURCE STRENGTH AT TIME AFTER RELEASE (Mev/cc/sec)
Mev/ Gamma 0.0 Hours 0.5 Hour 1 Hour 2 Hour 0.8 3.74 x 108 2.18 x 108 1.76 x 108 1.27 x 108 1.3 1.44 x 108 6.90 x 108 5.68 x 107 3.92 x 107 1.7 1.5 x 108 7.81 x 107 6.32 x 107 4.25 x 107 2.5 1.22 x 10e 5.97 x 107 4.76 x 107 3.30 x 107 4.0 3.17 x 108 3.24 x 106 2.34 x 106 1.27 x 106 5.0 9.01 x 106 6.81 x 104 6.74 x 104 5.48 x 106 6.4 2.86 x 104 3.61 x 102 o
o Mev/Gama 8 Hours 1 Day 1 Week 1 Month 0.8 5.83 x 107 3.10 x 107 8.38 x 106 2.16 x 106 1.3 1.61 x 107 3.74 x 106 4.26 x los 1,19 x jos 1.7 1.29 x 107 2.41 x 106 6.98 x 105 2.04 x 10s 2.5 7.04 x 106 3.56 x 10s 6.72 x 104 2.09 x 104 4.0 1.86 x 105 3.97 x 103 4.67 x 102 1.34 x 102 5.0 1.22'x 104 2.31 x 102 o
o 6.4 0
0 0
0 w
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I Table II.B.2-3 CONTAINMENT SUMP POST-ACCIDENT SOURCES
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GAMA ENERGY SOURCE STRENGTH AT TIME AFTER RELEASE (Mev/cc/sec)
Mev/ Gamma 0.0 Hours 0.5 Hour 1 Hour 2 Hours 3
0.8 9.17 x id' 7.19 x 10P 5.84 x id' 4.09 x 10' 1.3 3.23 x id9 2.35 x ids 1.98 x id9 1.43 x ld 9
l.7 2.00 x 109 1.88 x 109 1.67 x id' 1.25 x 10 2.5 1.16 x 109 7.16 x 108 5.84 x 108 3.60 x id8 7
7 4.0 6.79 x 10s 1.20 x los 8.21 x 10 4.48 x 10 5.0 3.18 x los 2.55 x 105 2.56 x 106 2.08 x id8 6.4 1.09 x 106 1.38 x 104 0
0 Mev/Ganna 8 Hours 1 Day 1 Week 1 Month 7
0.8 1.79 x 109 9.53 x 105 2.70 x 108 8.00 x 10 -
7 8
1.3 5.89 x los 1.41 x 108 1.62 x 10 4.54 x 10 7
8 1,7 3.84 x los 8.99 x 107 2.65 x 10 7.73 x 10 8
7.96 x 10s 2.5 8.13 x 107 1.01 x 107 2.55 x 10 4
3 4.0 6.79 x 106 1.48 x 105 1.78 x 10 5.08 x 10 5.0 4.63 x 10s 8.81 x 103 0
0 6.4 0
0 0
0
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l Definition of Radiation Map Zones After a LO.C.A.
Upper Dose Rate Umits(rem /ht)
Time After the Accident Occur.
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l l ,7, ya,ve Room t III I1 u 124._ ir2 aio, reed Recycie EvpI N"'" 3. ..s _ OA- ',;k j nuh 'I'*I ) ' 'l 17 9 128 ~ 127 173 Cnemcal Dran Tank Room L Q~ I { o LS l '74 Lauad'v Ha'd up Area 1 INI4e p'1 O i.!T' IZI" ) i'.
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I' i 5 [l 173 Component Cochng Water I F l ) I' jk d Heat E achanger Roc n .w j 5%I L [-] j uu
- 126,
,129 176 f ka Drain Tank Room 177 Floor Dean Tank Pump Room ~ k u e u e u n'%I j MjyIyy?IyIjy 1. II.! 178 Waste Mondor Tank Pump Room unuuns 18 0 3NININI Part6al Plan at EL 792'-0" 479 Corndor 180 Con dor i h 181 Recycle hold Up Tank Room k* " ~ ~ y 182 Ausdiary Steam Dram Tank Equipment Room -j 184 Waste Mondor Tank No 1 + Rmm IE n 185 Waste Mondor Tank No 2 u [ l O O O I ] ~ nam Partiel Plan at EL 792'-0" 11G Battery Room #2-2 ] 'C' ~ ~ 117 Banery Rorwn #1-2 ':= ~ m 118 UPS & Distnbution Rooms. (' 119 Laan 0 o u T]I[ O O __. O I i20 UPS & Distnbution nooms. r_ 121 Tra.n K b "3 122 Comdor v i23 canery Room #2-i us3 - n ri q. u3, 124 Battery Room #1-1 U U LJ 125 Cornaar O I F(~ O Q l w 126 UPS & Distnbution Rooms h--- g- [*=1 127 Banery Room #1-3 O u4 L JJ 128 aattery noom #2-3 000 129 T,a.n c-
- s. r,,
,a C 130.131 Stawways Plan at EL 778*-0" & 790'-8" Comanctie Peak S.E.S. Final Safety Analysis Report Post Accadent Dose Rate Map NOYES: Aumdiary & Elertncal Control Bkigs.
- 1. See Figure No.110 2-1 for ra+ahon zone legend Plans at EL 778 -0"& EL. 790'-6"
- 2. Radiation zone desegnanons for Ams. # 181. and ad acent 4 792'-0" i
areas represer t post-accident dose rates Fegtwe No il D 2-15 ~... w.. ) 2.. o . (. h +iil i Pian ai nevaimn. Sar-r & Sir-r j 'i 4 I 133 CaDe Room f l / 1J 0 88 2% 134 Catte Room 100 W 1344 Stair No FC-8 { // [b .h... l ll 134b Star No EC-2 / /' 7 } M M.. l [ 188 Floor Dram Wasse Evap. // P i , / / -- C_ Package Roons / ,\\ fffjf ff'"Iffff, 'J. 189 Waste Evaporator Room Ifffff'fMff'"^ 190 Wasie Evaporator Control e f' N // U Room / 207 l ll l 191 Spent Resan Skate Pump g,'e,,C ) f i / ,e ?"- Room to S'
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Partsal Plan at EL 822 -0* 192 Spent Rsaan Storage Tank f bp N J, g Room w 4 197 198 IJO $$ 4 5 - 193 Boron Recycle Evaporator d Room 1%f;,eU$%I<eeU%SN 195 Censntugal Chargrig Pump i -w~~~-f~~ U( 203 ia WS!!I209lllll"lf) W 8'0E Y, Ibs',$%,$ %%$,,%%%$N.%? m i ~~ m igs Pos,t,vo o,spim g-'
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.u 06 197 Componert Cochng Water Chargog Pump Room / n /. , yMyWp NN , /%% Partial Plan at EL FIT-F' Pump Room M , 05 y 198 Component Coohng Water I803ay $,,%I: eeff,e< f g99 Pos,t,ve Desp6acement i w + t n e. (g %9 Z%: O O Zy%% e Ctwgog Pump Room ' ' %Ih "f-- 3 II% % C....... _L-200 Centntugal Charging Pump 15 i QVs> kYn%% s ~R s %%i%%%%%%%%%Y '*"$' "****""M Roo
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$%%%:HHS%%S%%" 202 wve Room """"""'$c n
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203 wve noom j 208 Component Coolmg Water Pump Room 205 Component Coohng Water Pump Room 206 Borc Acx! Storage Tank Room E 20' Comdor D D Ll D D 207a RecycssHold-UpTankRoom 3 255 Filter Drop Area 134 gn Par 16al Plan at Elevacon 322'-O 208 Operatog Valve Hoosa O c I I O O Partial Plan at Elevallon 822'-0" 209 Operatog wve Room F 0 R n ' 17 0 l.-. I -dr ] = 1- ' r W i. Plan at EL 807-0" & 810'-6" Comanche Peak S E.S. Final Safety As.alysis Report Post-Accident Duse Rate Map NOTES:
- 1. See Figure No. Il B 2 -1 for radiahon zone legend Auusha;y & Electrcal Control Badgs.
- 2. Raaation zone designatens for Rms.189.193.199.200,201 Plans at EL 807-0"& EL 810'-6" 207a and ad acent areas rnpresent post accident dose rates Figure No. It B ?-16 i
e i 3. ~ ~ l I E f Pian at Esevatene s30%0 & s31%s - i 2l 210 Rm # Rm Name t 210 l #1 x 135 Controe Room l 'l MnJ
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h I t b WYl PIP a MM '3 W ,,1 I s 139 Production SupervWs ,j l $[ PPWFL*":g%"%Bdi : -. ' ML i h 13a Sta,r No EC~2 , f,f. 9 ,'e wn i( 'Ylf', A .. - y[. Wi d O ' 4 / f s. ,fy ) Othce i ; i.? (, a; h, d fjjj 1 140 Corrukw ff gli %; iO ... em ...x- 'a -=; s. - ' '~ $,.Gyf 'fy; 7 + ' 4,.. 's / 2,1 f: ,l ii! i o.i 141 Ba* ram g'a, ..i e .g ef, 142 Locker Room f
- 11. g+w.j! gfd yg, 143 Jarutor Closet J
St No EC-t g Partial Plan at EL 841*-0" 146 Charts and Supphes Ssorage $jMid j :M:h k k'- U 147 Correter Room ~ 207a Recycle Hold Up Tar
- Room 219 210 Piping Area nert
" " ~ ~ ~ " l C 218 Spent Fuel Pool Dominerahier e 7y Cubcle [ ar.- 212 Piping Area '3 4 Ma 213 Piping Area f 207 Y 214 Piping Area g i 323 224 N Pf, f Q h {L i 217 Denunerahiers Cubicles I 215 cen=neraniers Cubicies
- l 216 Pqwng Area llll r
,,,,~ ~$$II- ~ ~, ~ ng ,-~ em w c= bh 's$$$$$$$$$$$$$$$$$$$$s '#IIIIIIIIIISI$IIIIII%,, '$5$IIS - s compressor i l (No 1)Hi.wwn ] 's$$$?$$ffIff'fffff ?"" 221 Hydrogen Recombiner Room -g W 222 Wasta Gas Compressor (No.2 Roorrt [ 223 Wive 224 Hydt-Recombener Room 225 Bors 'id Tank Room O O O D D 226 Corrdor Agr nr} 7-pm-- uui u Partial Plan at Elevation 841%0" r i 227 Valve Operating Roorn ~ 13 5 hm m# 145 3 i-a may,,-, a W esee:s & i i CLTJU 139 141 84 3 tiTITJ 136 137 14 6 14 7 3 ie , 14 2 14 4 e Plan at EL 830'-0" & 831'-6* T* Comanche Peak S.E.S. Final Safety Analysis Report 130TES: Post Acculent Dose Rate Map i Seo Figure No il B 2-1 for radiation zone legend Auudirry & ElecincalCont,ce Bksgs
- 2. Radiatum zone designatens for Ams. # 207a 213. 215,216. 217. 218. 220,221. 222.
Plans at EL 830%0"& EL. 831*-6" 223.224 and a.1 acent areas represent post-accadent dose rates Fafne No il B 2-17 .e M& !)i' e ) o CS w T( ~ se s k r t n o n t s S n n e r p t o a C oa c r r pM e ~ s o o t o o iv m m p S. F te S r R R o l d w mm=Nbw*cz t a o lo a lot p E.i R a 2p s n 1 b
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- ^'*****
%:' % % % % %,%,1,n u,%,% % % % %,e'%,%,~% % % % % %,% %,,:,~ ~s i, %%%,unnn ' :' H: g 240 Gas Decas Tarss. n in~~~~ ~ isisii~ n n1 Co'"o'wm** e,% % % % % % % % % % % '% ' % '% % % % % % :):'%: % P 4,,%Mg%%%SS% %%%"'"%$ $'W'"s"S"'S'"_$,%,,%,,: 241 con - i y ', u 242 Gas Decay Taras Drain h,j / ' / Partial Plan at Elevation 862'-8* f, i nJ l mn=- 243 ./ 239 r, y I 243 Vatve Operatmg Room h hh kh ?4 1 P, S /. ,$ 'i,$ $, h',$ ' i$ $b h
- #*~
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s=,d ~l O A h 151a ~ g 151b '] m q g 'l Plans at EL 852'- 5" & 854*-4" Comanche Peak S.E.S. Final Safety Analysis Report Post-Acculent Dose Rate M@ Nns at R 8W,Mr@ps. Auxmary & Ntncal agure No 118 2-1 for radiation tone legend & EL 8M -4
- 2. Radtahon zone deswjna. ions for Ams # 234 235,238. 239. 2 to, 242,243 and adpcent areas represent post accident dcse rates Fupwe No. Il 0 2-20 J
~* e eg.g. s Plan at Elevation 81F-8* h,:uba,22,32/>>>sh;;;,;;;;;;;;;;;;;;;;;;; Rm a Rm Name
- =::::::::::==:=::::==::::::
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248 Radrodd Loaeng and l lllfjlllllllll:flfff'f n Unnoaaing ' f =S ::::: f :::::::::: ::::::S,I:fif '3 * ~- l "'*^ s ,fu 249 Pump Room W Q Ws;'a ^ 250 ' ' a fffisss 249a PumpRoom I'fffff$fl' 'S$f),'; 'I$fiff fIff$) O.(l// I l 'fIIII 2490 Open Area 'Ik49'f. Ilf4'9al I I' I 250 Decontamwsabon Area c g~"' th'fffffII) ~ .'s gr; % ; ff:fffSi 2*' 1 % ";; igjf;; ifSfS) f'.2%s, r '. 1 250a space a "fu l'"f' ?fY, y 'jfffI isa 'f s ff I 2500 Change Room us
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'e/ Mf; f 3 f' 251 Drum Stora'3e Area f,,5lf,};"iff,$f,%"S,SS )$SSSS?@ a u ____16 cit, 254 Stan No. F-1 f,$f$SA 'fff ,P,4 af ' 249b""' 247e e-252 DrumFdisng Area I!fd.f f,,,$, f a 253 Drum Slotage Vewsng Area f'ffffff IIIifS' n 255 Open Area </ssa ,iss .Kf3%$$$ fffIIJ '250al b'fyy ffffII fffff ffffff! IIII' m @ns W Roorn ?$$$ Partial Plan at EL 825'-2"& M2'-6 l ffffI) d 'ffffff; (MN b. $$y' 257 Low Levet Raeation Storage m fffflI 'f,'ff M f" a $$/ 257a Open Area "ff' Mil' I \\ f', y' ua F "I 258 Wet Cash toadng Area M [b-i 'I i fr 3. 259 Spent Fuel Pool No 1 f r,f,fffff-L f, f 4.- k 260 Spent Fuel Pool No. 2 K4 san ( J 57 15ff ffffffI Partial Plan at Ekvatione 825'-2"& 't i fffffg
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==a 247s Tunnes / Plan at EL 810'-6* 24Ta l - pPartiel Plen at EL 802'-0" w Comanche Prok S.E.S. NOTES: Final Safe;y Analysis Report
- 1. See Figure No. II 8.2-1 for radsation zone legend Post Ar.adent Dose Rate Map
- 2. Radiation zone dessgnations for Rms. 251,253.253 & 260 represent Fuel Buelding norrrwi operation dose rates Plans at FL 810'-6", EL 825'-2"
- 3. Radiation zone designations for Rrns. 252,261 & 262 represent dose
& EL 802'-0" rates dunn0 waste solidification process system operation Fagure No. Il B 2-21
e e st, ' ',.i e +,a Man at Elevatxans 838'-9" & ip41*- Er p +- O>>Ve>>>V,Ve>V,>V >>>yA RM # Rm Name
- Illll$l':lllllllllllllll 258 wee Cask Loanen Area 259 kweni M M % 1 lII'rII/I,'IIIIIII'IIIIIIII,
/ 269 260 Spent Fuen Pace No 2 sum, nimimsini l llllly',llllllllllllllllll 263 Open Space yllllllllllllllll 263a P4 e/ </// sum, sin /sn us m n n u m 264 Spen l Fuel Pool Heat
- ' IlII ll
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- IIIIIIII'
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- I(2'It/
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- 2. Hadulion ior,1esonatens for rooms 259. 260 & 270a represent Plans at EL 838'-9"& EL. 84t'-0" normal opera. m dose rates.
Figure No il D 2-22 ~ 6_ + + = * * + =.--< +e we. %w.%-
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