ML20092A418

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Proposed TS Tables 2.2.1-1,2.2.1-1,3.3.1-1,3.3.1-2, 4.3.1.1-1,3.3.2-1 & 3.3.2-3 Re Main Steam Line Isolation & Automatic Reactor Shutdown Functions of Main Steam Line Radiation Monitor
ML20092A418
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 02/03/1992
From:
Public Service Enterprise Group
To:
Shared Package
ML20092A414 List:
References
NUDOCS 9202100104
Download: ML20092A418 (19)
v • d • e


Text

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IABit 2.2.1-1 5n REACIOR PR0llCil0N SYSIIM INSIRUMENIA110N SEIPOINIS

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7. Drywell Pressure - High 5 1.68 psig 1 1.88 psig
8. Scram Discharge Volume Water Level - liigh
a. f loat Switch L levation 110' . J J.5" Elevation lil' 0 S"
b. Level Iransmitter/Irip Unit Llevation 110' 10.5"* Elevation 111' 4 S"a
9. Turbine Stop Valve - Closure $ 5% closed s /'A closed 7
10. Turbine Control Valve f ast Closure, Irip Oil Pressure - Low > 530 psig > 465 psig
11. Reactor Mode Switch Shutdown Position NA NA
12. Manual Scram NA NA "80.5" above instrument zero EL 104' 2" for Level Iransmitter/Irip Unit A&B (South Header) 83.25" above instrument zero EL 103' 11.25" or t vel Transmitter /Tr' W

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LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

4. Reactor Vessel Water Level-Low The reactor vessel water level trip setpoint has been used in transient analyses dealing with coolant inventory decrease. The scram setting was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the fuel and pressure limits.
5. Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events. The MSIV' e closed au_tomati all from measured parameters such as high steam flow, .,, ..q: ~ "n6_ Fadicticq, low reactor water level, high steam t. uel l

temperature, and Tow steam line pr ssure. The MSIV's closure scram an( cipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal / hydraulic Safety Limits.

6. -Mein Stccr Line Red 4+t4ee-44eh-DEIRI' 1 The mai :t00: 'ine r-adiat4on-detecton-we-prw4deddo-detect-a-grou fahe of the fuel cladding-Wheft-the high radiation is deteetedr e-trip-+s

-iei-t4*ted-to-reduce-the-centfeued-f ailure of-feel-cladding. At'-the same t4me-the-maifHrtcam line i; clot 4en-valves-are cle;cd-to44mit-the relcate of 'iction prcducts. 'hc trip Octting i; high encugh abcvc backgrcund radiction level; 10 prevent spurious trips yet icw enough-to-prompt 4y-detect-alrow-f444ures4n-them fuel cladding, [ ~ ^ -

1- --

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7. Drywell Pressure-High High pressure in the drywell could indicate a break in the primary pressure boundary systems or a loss of drywell cooling. The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy l being added to the coolant and the primary containment, The trip setting was l selected as low as possible without causing spurious trips.
8. Scram. Discharge Volume Water Level-High The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram. Should this volume fill up to a point where there is insufficient volume to accept the displaced water

, at pressures below 65 psig, control rod insertion would be hindered. The reac-l tor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough l

to accommodate the water from the movement of the rods at pressures below l 65 psig when they are trinned. The trip setpoint for each scram discharge l volume is equivalent to a tantained volume of approximately 35 gallons of water.

HOPE CREEK B 2-8 l

l-1

TABLE 3.3.1-1 (Continued) .~

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m REACTOR PROTECTION SYSTEM INSTRUMENTATION n

x .

  • APPLICABLE MINIMUM OPERATIONAL OPERABLE CilANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a) ACTION

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7. Drywell Pressure - liigh 1, 2(h) 2 1
8. Scram Discharge Volume Water Level - liigh
a. Float Switch 1, 2(g) 2 1 5 2 3 u

Y

b. Level Transmitter / Trip Unit 1, 2(g) 2 1 5 2 3
9. Turbine Stop Valve - Closure I Id) 4(k) 6
10. Turbine Control Valve Fast Closure, Valve Trip System Oil Pressure - Low 1(3) 2 5) 6
11. Reactor Mode Switch Shutdown Position 1, 2 2 1 3, 4 2 7 5 2 3
12. Manual Scram 1, 2 2 1 3, 4 2 8 5 2 9

S TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION ACTION 1 -

Be in at least HOT SHUTOOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 2 -

Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within one hour.

ACTION 3 -

Susper.d all operations involving CORE ALTERATION 5* and insert all insertable control rods within one hour, ACTION 4 -

Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ,

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ACTION 5 -

ve in STARIUP ith the main :-team line isolat4e+-v*hes-cawed. !

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f ACTION 6 -

1 i tea a reduction in THERMAL POWER within 15 minutes and rw ace turbine first stage pressure to less than the automatic bypass setpoint within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

ACTION 7 -

Verify all insertable control rods to be inserted within one hour.

ACTION 8 -

Lock the reactor mode switch in the Shutdown position within one hour.

ACTION 9 -

Suspend all operations involving CORE ALTERATION 5*, and insert all insertable control rods and lock the reactor mode switch in the SHUTOOWN position within one hour.

'Except replacement of LPRM strings provided SRM instrumentation is OPERABLE per SpeciJication 3.9.2.

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HOPE CREEK 3/4 3-4

7 TABLE 3.3.1-2 -

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REACTOR PROTECTION SYSTEM RESPONSE TIMES S INSERP 1 m

^

,/'

/ RESPONSE TIME

/

/ FUNCTIONAL UNIT (Seconds)

1. Intermediate Range Monitors:
a. Neutron Flux - High NA
b. Inoperative NA
2. Average Power Range Monitor *:
a. Neutron Flux - Upscale, Setdown NA
b. Flow Biased Simulated Thermal Power - Upscale 5 0.09**
c. Fixed Neutron Flux - Upscale 5 0.09
d. Inoperative NA R 3. Reactor Vessel Steam Dome Pressure - High 5 0.55

MM Ne M ure Yadiatier

- High - MiaC Hi'h '

~rywell Pren

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7. A
8. Scram Discharge Volume Water Level - High NA
a. Float Switch NA
b. Level Transmitter / Trip Unit NA
9. Turbine Stop Valve - Closure 1 0.06
10. Turbine Control Valve Fast Closure, Trip Oil Pressure -. Low . 5 0.08#
11. Reactor Mode Switch Shutdown Position NA
12. Manual Scras NA g
  • Neutron detectors are exempt from response time testing. Response time shall be measured g from the detector output or from the input of the first electronic component in the channel.

g **Not including simulated thermal power time constant, 6 1 0.6 seconds.

r* # Measured from start of turbine control valve fast closure.

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2 TABLE 4.3.1.1-1 ,

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E REAC10R PROVECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS' "o

CHANNEL OPERATIONAL Q

  • CHANNEL FUNCTIONAL

' CHANNEL CONDITIONS FOR milch

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FUNCTIONAL UNIT CHECK TEST CALIBRATION (#} SURVEILLANCE REQUIRED

1. . Intermediate Range' Monitors:
a. Neutron Flux - High S/U(b) 5.

S/U(c) ,W R 2 S W R 3,4,5

b. Inoperative NA W NA 2,3,4,5
2. Average Power Range MonitorII}:
a. Neutron Flux - 5/U(b) 5, S/U(c) ,W SA 2 Upscale, Setdown S W SA 3, 4, 5 w b. Flow Biased Simulated A Thermal Power - Upscale S,D I9) S/U(C) ,Q W Id)I') SA,R(h) y w

0 c. Fixed Neutron Flux -

, _ Upscale 5 S/U(c),Q W(d), SA 1 L

d. Inoperative NA Q NA 1,2,3,4,5.
2. Reactor Vessel Steam Dome ~

High 5 Q(k) R 1, 2

' Pressure

4. Reactor Vessel Water Level - 1, 2 Low, Level 3 5 Q(k) R v

g 5. Main Steam Line Isolation 1 NA- R e Valve - Closure- Q 7 E Nin St = Line Radiati= - N 6.

5 _44 ! gh , utgh-.- S Q-R- -1. MN  ; ,

z P 7. Drywell 1, 2 Pressure - High. S Q(k) R INSERP 1 l w- n

IASLE 4.3.1.1-1 (Continued) #

t REACIOR P90iECI10N SYSIEN IIISIAISENIAI10Il SURVEILLANCE REQUIREM[NIS .

n CHA00EL' E CHAINIEL fisICIIGIIAL

' OrtRAI101 SAL .

E CHAINIfL ftSICil0114t Ull!I CHECK IEST CON 0ill0NS 10R WHICH CALISRATION SURVfIttAfect RtQUItie 8.. Scran Disc W Water Level - Mi

a. Ilaat Switch NA q R 1, 2, SIlI
b. Level Transmitter / Trip Unit S .q"I R 1, S Ill
9. Iurbine Step Valve - Closure IIA q R I j
10. 'larbine Centrol Valve Fast Closure Valve Trip Systes  !

Gil Pressure - Low M4 q R I

11. Reacter flode Switch O Shutdown Position IIA
  • R ItA 1,2.3,4.5  !

w 12. Manual Scram IIA W NA e's 1,2,3.4,5 '

(a) Heutron detectors may be excluded from CHAleEt CAtl8AAll001. ~

(b) 1he IM and SM channels shall be determined to everlap for at least 4 decades during each startup  !

af ter entering OPERATICIIAL COWillell 2 and the INI and APWI channels shall be determined to overlap i for at least.% decades during each controlled shutdenne, if not pertensc<f within the previous I days.

(c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startage, if not performed within the previous 7 days.  ;

(d) This calibration shall consist of the adjustmen;. of the APWI channel to confers to the power values

~

4 calculated by a heat balance during SPERATICIIAL COINITI0ll 1 when IHEMIAL POWR > 25% of RAllD

{ llEWIAL PRER. Adjust the APWI channel if the absolute difference is greater than 2% of RAl[D IHil#tAl

+

s POWER. Any APWI channel gain adjustment made in compliance with Specification 3.2.2 shall not be j included in deteretning the absolute difference.

a (e) 1his calibration shall consist of the adjustment of the APWI flow biased channel to confors to a calibrated flow signal.

IT  !'

  • (f)' The LPWis shall be calibrated at least once per 1000 ef fective full power hours (If Pil) esing the 11P system.

(g) Verify measured core flew (total core flow) to be greater than er equal to established core flow at the existing' recirculation loop flew (APAN I flow). ,

(h) This calibration shall consist er y t

( i) ":iiO_-OY2 W .-= C Md 'l g the 6 t_ 0_.6 second simulated thermal power t ea.e c oast arit . A  :

g ,,,, 1 ;; ;. _ _ - v=__t ' Sithht
OMiri'FJE19

~ _ _ -

-Et :-G-N2 (j)  : any control r withdrawn. Not applicable to control rods removed per spec s t iut ion 19 to I or 3.9.10.2.

(k) Verify the tripset point of the trip unit at lea.st oene per 92 davs

, a O IABLE 3.3.2-1 (Continued)

A n IsotATION AC10AY10N IN51RUMENTAT10N -

E 9 VALVE ACIDA-110N GROUPS MININUM APPLICABLE IRIP FUNCIlON OPERATEkdgY OPERATIONAL SIGNAL OPERABLECHANNEQ)

PER 1 RIP SYSitM CONDill0N ACTION

3. MAIN STEAM LINE ISOLATION
a. Reactor Vessel Water Level - 1 2 1,2,3 Low Low Low, Level 1 21
b. h in Steam Line Radiation - 1, 2(b) 2 1, 2, 3##

c.

High, High h in Steam Line Pressure - 1 2 21

]

1 22 Low DctRr 3

d. Main Steam Line Flow - High 1 2/line 1, 2, 3 20

$ e. Condenser Vacuus - Low 1 2 1, 2**, 3"* 21

(;> f. Main Steam Line Tunnel g Temperature.- High 1 2/line 1, 2, 3 21

g. Manual Initiation 1, 2, 17 2 1,2,3 25
4. REACTOR WATER CLEANUP SYSTEM ISOLATION 1/ ValveI *I
a. RWCU a Flow - High 7 1,2,3 23
b. RWCU a Flow - High Timer 7 1/ ValveI 'I 1,2,3 23
c. RWCU Area Temperature - High 7 6/ ValveI 'I 1, 2, 3 23
d. RWCU Area Ventilation a 7 6/ ValveI ') 1, 2, 3 23 Temperature-High

" e. SLCS Initiation 7 III 1/ ValveI ") 1, 2, 5# 23

.E Reactor Vessel Water 2/ ValveI '}

f.

7 Level - Low Low, Level 2 1,2,3 23

g. Manual Initiation 7 1/ ValveI 'I 1, 2, 3 2'-

e TABLE 3.3.2-1 (Continued) .

ISOLATION ACTUATION INSTRUMENTATION TABLE NOTATION A

me l VALVES CLOSED BY SIGNAL TRIP FUNCTION

3. MAIN STEAM LINE ISOLATION 1 (HV-F022A, B, C & D, HV-F028A, 8. C & D, HV-f06?A, 8,
a. Reactor Vessel Water Level - C & D, HV-F616, HV-F019)

Low Low Low, Level 1 ')

b. Main Steam Line Radiation - High, liigh 1 (as abov }2
c. Main Steam Line Presure - Low 1 (as above) -
d. Main Steam Line Flow - High 1 (as above)
    • e. Condenser Vacuum - Low 1 (as above) w Main Steam Line Tunnel 1 (as above)

[g f.

Temperature - liigh 1 (as above), 2, 17 (SV-J004A-1, 2, 3, 4 & 5)

g. Manual Initiation
4. REACTOR WATER CLEANUP SYSTEM ISOLATION 7
a. RWCU a Flow - liigh 7
b. RWCU A Flow - High, Timer 7
c. RWCU Area Temperature - High

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23:waar ::N*a!%=ENT 150L AT!04 4.

teacice vessel s'ater Level

1) Lew Low, Level 2 NA
2) Low Low Low, Level 1 NA
3. Orywell 8eessure - Nign
c. NA Reactor Builcing Exnaust Raciation - Hign
3. NA Manual *nftiation MA 2,

SECON0aRY CONTAINMENT ISOLATION

a. Reactor vessel Water Level Low Low, Level 2
3. NA Orywell Pressure - Hign MA
c. Refueling Floor Exnaust Raciation -

< 4.0 Nign50)

c. Reactor Building Exhaust

< 4.0

~

Raetation HighID)

e. M4nual Initiation NA
3. MAIN STEAM LINE ISOLATION a.

Reactor vessel Water Level - Low Low Low, Level 1 D. ( 1.0*/c Main Steam Line Radiation High, Migh(a)(b) g 2 1;_7 _47-

c. Main Steam Line Pressure - Low
d. Main Steam Line Flow-Nign Y I.0* R 13(4'l.,
e. fondenser'Vacuus - Low 7 0.5'/7 13 *)

~

f. Main Steam Line Tunnel Temperature - Nign RA MA
g. Manual Initiation MA y4 4

REACTot WATER CLEANUP 5YSTEM !50LAT!0N

a. RWCU A Flow - Mign NA
e. RwCU a Flow - Hign. Timer NA c.

RWCU Area Temperature - High NA

c. RWCU Area Ventilation A Temperature High , NA
e. SLCS Initiation NA
f. Reactor Yessel Water Level - Low Low, Level 2 NA
g. Manual Initiation MA 5.

4EACTot CORE ISOLATION COOLING SYSTEM ISOLATION

a. RCIC Steae Line a Pressure (Flow) - Hign '

NA

b. RCIC Steen Line a Pressure (Flow) - Hign, Tirar NA
c. RCIC Staam Supply Pressure - Low NA l 3. ACIC Turnine Exhaust Diaphragm Pressure - Hign MA HOPE CREEK 3/4 3 26 l

l

Ref: LG 89-13 4

ATrK3MFNP 3 OcNPIJMG WJ3){ NRC CtNDITICNS FW REFURIM3__HilD 314LXB

. = . - . . - - , - - - - . . . . . - . -

hTINAtilME.)

coNorricus ,

%e NRC staff concluded that the remval of the HSUM trips that autcentically shut dcun the reactor and close the MSIVs is acceptable ard that the Licensirg Topical Report, NEEO 31400A, could be referurccd in support of our amedmnt request provided that:

1. Se assunptions with regard to input values mde in the gercric amlysis of the LTR are bourdity for the plant. . .

%ble 1 of this attachnet prinidm a emparinan of key irput puamters and

%blm 2a, and 2b ocuparu dose assesamt betweart the licpe Crunk Ccreratirg Station (IKDS) UlGAR ard NEID 31400h amlysis assunptiam.

2. Reasomble assurance is provided that significantly ircreased levels of radioactivity in the min steam lines will be controlled expeditiously to limit both occupational ard environnental releases. . .

IKUS has, in place, procedurm that ensure that any significant ircruase in the levels of ItElloactivity in the main stem lims is prmptly ccritrolled to limit erwiumwazd.al releasm ard on-site (occapaticml) mmr=3rus. 2000 procodtrrm have been reviewed ard wi.ll be :pgraded, upan recedpt of the rugC%d awamht, to ensuru their contiruod applicability ard cxarths.

3. %e MSUN ard offgas radiation monitor setpoints are stardardized at 1.5 tims the nitixgen-16-backgreurd dose rate at the monitor locations ard should either or both exceed their alam setpoint, the reactor coolant will be prmptly sarpled to detemine activity levels ard the possible need for additional corrective actions...

'1he IEIH4 retpoint is 1.5 t inm the N 6 background at the monitor location.

'Ihat alarm would trigrpr entry into the abnormal m( ktre, OP-AB.ZZ-203, i

which requirm a rmctor coolant sanple be @tained ard analyzed. %e Offgas Rasilation Mcnitor alarm is set to satisfy ll0CE 'IS 4.11.2.7.2.b by alarmirg at 50% frcrease (1.5 timm)* the naminal stely-state fission gas rulmse frun the reactor coolant, after factoring out any incruases due to changes in thermal power leve Ixpresentative gas sanple taken frun rear the clischarge of the main condenser air ejector ard wculd trigger mitry into me or more of the above almannal procedums - which, in turn, prescribe further additional corructive acticm.

1 1

Attadarit 3, (cuit'd)

  • 'Ihe of fgas pre-treatment radiation ronitor alarm is set at 1.5 tires backgruind or 10 mr/hr, whichever is gmiter. 'Ihis 10 mr/hr caveat has bem fourd necessary to elimimte r:arerous spuricus alarrs (with their attendant distractions of the control roam operators) due to current background levels so Icv (4 to 5 mr/hr) that cirulit noise or minor charycs in offgas flcurate can initiate an alam. 'Ihe 10 mr/hr alarm setpoint cormspords to .05% of the limit of 330 millicuries /secord specified in TS 3.11.2.7. It is in accordance with this 'IS that the offgas radiation monitor alarm is cet.

Historically, as a point of reference, one leakiry fuel pin has prcductd several thousand mr/hr levels on the offgas radiation monitor at 1101S.

Themfore, the currrnt alam set point of 10 mr/hr provides conservative indication. As backgreurd levels innuase with plant age, the 10 mr/hr alam will eventually be supplantrd by the 1.5 tires backgrourd alam setpoint.

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

  1. t T. -

' TAIRE 1

'. . CCMI%RISCW OF LEY ANALYSIS INIUP VAIUF3 m IKE.UPTM _VS. NFD2_314QQh pg AME3TR NEDO 31400A VALUE (*) llOGS UlTAR VAIUE  !

No. of failed fuel rods 850 770 Cbru average power (MWt)- 3579 3458 (305%)

Relative power level of 1.5 (rano) failed rods (fraction)=

Power level of failed rods (MWt) 0.12 0.11  ;

Fission Product (FP) _ release - ,

-from failed rods Melto:1 100% NG / 50% Iodines (same)

Non-Melted 10% IC / 10% Iodines (same)

Mss fraction of melted fuel 0.0077 [sano)

% of ?? transported to Main 100% NG /10% lolines (same) '

Condenser

% airborne of FP in Main 100% NG /10% Iodines (same)

Corrienser For CRDA without MSIV isolation,100% of the Noble Cases (10) are held-up in the Off-Gas Treatment systan charcoal beds for a time; the Iodines are retained indefinitely in the clarcoal beds.

- min Condenser leakage . (f) 1% per day (same]

Off-Gas Treatment System 112 Recambiner/ Charcoal Treatment Systan HOGS. specific (#f) (322,000 lbs charroal)

IELBYIBSS of chartml is nossible Charcoal bed holdup times:

(65 F/40 F dewpoint) Kr = 35.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Xe = 34.3 days ,

(77 F/45 P dowpoint)(**) Kr = 20.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> Xe = 15.3 days H -flow rate to recambiner - 50-150scfm 154 ccfm

( ign capability)

. Air / Noble' Gas < flow rate site specific 75 scfm

. 'Ihyroid dose conversion factor- Reg. Guide 1.109 .(same)

- Breathirg Rates : Reg.Cuide 1.3 (same)

Whole Body Dose Conversion Factor Reg. Guide 1.109 (same)

(semi-infinite cloud)

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

___._._._m....

'4 8

'INEE .1. , Continued t

i PAB6tg3E8 - NEDO 31400A VAIDE f*) llCro Uh$

Radiological Consequences 00taCD3 ' cot &OO1- I i

_ Evaluation (***)

Dispersion coef ficient, - X/Q 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Site 9oundary '(f f f)' _1.9E-04 (soc /m3) _

8 - 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> IPZ- (#ff)- '4 0E 05 (sec/m3)

m. -

L(*) Dccept as noted in- (f): and (f f)- below, values apply to the CRDA both y_ith MSIV isolation and without MSIV isolation.

(f). Applies only to CRIA y.ith MSIV isolation. a (f f)' Amlies cnly to CRIA without MIV isolation ard 100t, of Noble Gas .

- source term prrv,wM through the Off-Gas Treatment System. ,

-(**)- NURB3 0016, Rev 1 values

-(***) NEDO 31400A calculates the radiological consequences of a CRIA usirg ,

the CONACD3 code while the HCro UFSAR uses the earlier 00NA001 cxxk3.

GE mmo, -IRR-89-07,. dated S/9/89, has provided fuel activity rulcase fractions required to update the IKrs UTSAR to the Cot &OO3 code.

(##f)~ NEDO 31400A uses boundirg values of 2.5E-03 for CRIA analysis per SRP - -

- analysis ard 3.0E-04 for CRIA W/o EIV isolation. Dose calculations-are done for the HOGS-sps a cific X/Q values only.

?

f y v g ., ---w-e- W--? y- g w- - y- V ye-. ,pw

TAIRE 2a CRDA DOGE CIMPARISCH IERUllTMLYliM714006 bBhkYSLS_MQ10D 2_lMRuin1LIRMPFt inils flRM M DIE IDDY,  % (*) 'nNROID  % (*)

Present Design Ihsis 3.11E-02 0.50 2.62E-01 0.35 HCGS UESAR 15.4.9 ( f)

NEID 31400A Design Ihsis 2.50E-02 0.42 3.50E-01 0.47 (I ard it)

NEDO 31400A - with 10 MSIV ISO 1ATION ( **, i f , ard ** *)

(Charcoal Bcd 'Ibmperatures) 0 65 F 2.03E-02 0.34 N/A N/A 77 F 3.50E-01 5.80 N/A N/A Footnotes;

(*) Ibrrent of 25% of 10CFR100 (or 6 RD4 WB & 75 RD4 '1hyroid)

(1) Design Basis is MSIV isolation w/ Noble Gas & Iodine leakage fram Main Cbrdenser.

(**) 10 MSIV isolation with 100% of Noble Gas ptwM by Offgas Treatment System ard all Iodine retained irdefinitely in Charcoal Beds.

(##) HCGS-specific values used per Table 1 of this Attachment

(***) Krypton & Xenon docce obtaincd separately frun Figures 3 & 4 of NEDO 31400A ard given below. Whole Body dcse is the sum of the Kr ard Xe doses.

BED TEMP. Xe DOSE (RD4) Kr DOSE (RD4) 65 F 1.40E-02 0.63E-02 77 F 1.90E-01 1.60E-01 Doses Wre obtained frun NEDO figures at a X/Q of 3.0E-04 ard scaled to a X/Q of 1.9E-04 (nultiplied by 0.633) to eliminate interpolation.

. .. .,. . . . . , . ~ . .. - .. .. . - . . - - . . - . - - . . ~ . . . . - ~ . . . ,

. . ~ - - -

4 '

-5.. .e . 6

'INRE ab OEA DOSE CDG%RIfEM HOGS UISAR VS. NEED 314QQh NGOfSIS IE'Il0D 24 If1R Iai IGUIATM ZO4E DOSHS fRIM) *

}iDIE BODY  % (*) 'IinPOID  % - (*)

Present Design Basis 6.55E-03 - 0.11 5.52E-02 0.02

- IKES UlTaAR 15.4.9 NEDO 31400A Design Basis 5.26E-03 0.09 7.37E-02 0.03 NEDO 31400A-- with 10161V ,

'ISOIATIG1 .

(Chan:oal Bed TW'peratures]

65 F- 4.27E-03 0.07 N/A N/A-77 F 7.37E-02 1.20 N/A N/A l Fem:

(*) Percent of 25% of 10CFkl00 (or 6 RIM WB & 75 REM 'Ihyroid) e f

g .+1: - - ,a -,e ,,-w

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