Text

Rev 0 to Core Damage Estimate Procedure CH-TI.22-011(Q), Estimation of Reactor Core Damage Under Accident Conditions
	ML20127J356
Person / Time
Site:	Hope Creek
Issue date:	06/18/1985
From:	Moose T; Public Service Enterprise Group
To:	;
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References
	CH-TI.22-011(Q), CH-TI.22-11(Q), NUDOCS 8506260572
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(

!! OPE CREEK GENERATING STATION CH-TI.22-011(Q) l-ESTIMATION OF REACTOR CORE DAMAGE UNDER ACCIDENT CONDITIONS i

Prepared By:

Tommy Moos _e INEC) w Nem.

5/6/85 Date Reviewed By:

b[b Chemistry Engineer Date ALARA Review:

[I s AL 6'b d (

Radiation E~otection De t.

Date r

i Reviewed By:

'I Okk 4 F[

Site Er ineerin De p t./

Date 8

J

.(M

[-d'[I

~

Reviewed By:

j SQAE e

Date SORC Review:

0 Yli e /

Sf/9'fES 0 5~ #ll I

Chgtrman Date Mtg. No.

4 -/f-8'[

Approved By:

d. It g

J Teghnical Manager Date g506260572850624 l

R ADOCK 05000354 l

E l

PDR CH-TI.22-011(0)

Rev. 0

. ~ _.

ESTIMATION OF REACTOR CORE DAMAGE UNDER ACCIDENT CONDITIONS TABLE OF CONTENTS SECTION TITLE PAGE 1.0 PURPOSE..........................................

2 2.0 PREREQUISITES....................................

2 3.0 PRECAUTIONS AND LIMITATIONS......................

2 4.0 EQUIPMENT REQUIRED...............................

3 4.1 Apparatus...................................

3 4.2 Chemicals...................................

3 5.0 PROCEDURE........................................

3 5.1 Standardization.............................

3 5.2 Method......................................

3 5.3 Calculations................................

7 5.4 Acceptance Criteria.........................

7 5.5 Re p o r t i n g Re s u l t s...........................

8 6.0 ATTACHMENTS......................................

8

7.0 REFERENCES

9 CH-TI.ZZ-Oll(Q) 1 Rev. O

't-T ESTIMATION OF REACTOR CORE DAMAGE UNDER ACCIDENT CONDITIONS 1.0 PURPOSE 1.1 The purposes of this procedure is to provide an estimate of the reactor core damage from the measured fission product concentration and other confirmatory plant parameters under accident conditions.

The primary interest is to provide differentiation between no fuel damage, cladding failures, fuel overheating and core melt.

This procedure is derived from the NRC -

approved generic procedure prepared by the BWR Owners' Group.

r 2.0 PREREQUISITIES 2.1 The post-accident analyses shall be performed IAW

+

CH-EO.SH-004( O).

See steps 5.2.1 and 5.2.2 for the sample sources required during certain accident event types.

2.2 0 shows an example of the sequence of steps to assess the extent of damage to the core.

Subsections in this procedure may be performed in any sequence due to the variables and event types involved.

2.3 Personnel required to perform calculations IAW this procedure have been qualified IAW SA-AP.ZZ-014(O).

2.4 Verify that sample decay and pressure - temperature corrections have been performed IAW CH-RC.22-007(O).

3.0 PRECAUTIONS AND LIMITATIONS 3.1 The performed estimates are calculated under the presumption that no reactor coolant cleanup systems are operated after the accident.

3.2 Measurement of Cs-137 and Kr-85 activities may not be possible until the reactor has been shut down for several weeks to allow the decay of shorter lived isotopes.

3.3 For the purpose of this procedure cladding failure below 1% is considered a non-accident condition.

3.4 Be aware of Iodine and Cesium spiking that may result in erroneous results.

CH-TI.22-Oll(O) 2 Rev. O

4.0.

EOU !i'M ENT REOUIRED 4.1 Apparatus Not Applicable 4.2 Chemicals a

Not Applicable 5.0 PROCEDURE 5.1 Standardization Not Applicable 5.2 Method 5.2.1 Selection of Gas Sample Source Event Type Sample Location Non-Breaks (e.g.,

MSIV Torus Atmosphere Closure)

Small Breaks

1) Drywell Atmosphere (before depressurization) 2)

Torus Atmosphere (after depressurization)

Large Breaks in drywell Drywell Atmosphere Large Breaks outside Torus Atmosphere drywell 5.2.2 Selection of Liquid Sample Source Event Type Sample Location Pressurized Reactor Jet Pump Instrument Lines Depressurized Reactor RHR Loop in cooling mode 5.2.3 Determination of Core Damage from Fission Products Concentration 5.2.3.1 Obtain liquid and gas samples as indicated in section 5.2.1 or 5.2.2 from the post-accident sample station

{

IAW CH-EO.SH-001(O).

[

l l

CH-TI.ZZ-Oll(0) 3 Rev. O

~

5.3.3.2 Obtain the fission products concentrations in the samples IAU L

Cil-EO.SH-004(0).

Record concentrations of I-131, Cs-137, Xe-133 and Kr-85 and the time of sampling on Attachment 1.

NOTE 5.2.3.3 Decay correction will not be applicable if the gamma spectrometer performs this function.

5.2.3.3 Correct the measured concentrations for decay to the time of reactor shutdown as per Attachment 1.

NOTE 5.2.3.4 I-131 and Xe-133 need only calculations for 6 half-lives prior to shutdown.

5.2.3.4 Calculate the fission products inventory correction factors as per.

5.2.3.5 Calculate the drywell or torus gas volume correction factor (f ) and the g

primary coolant or torus mass correction factor (Fw) as per 2 if the liquid or gas masses or volumes are different than those used for Attachments 4-9.

5.2.3.6 By using the correction factors, calculate the normalized concentrations of the fission products as per.

5.2.3.7 Use Attachment 4 through 9 to estimate the extent of fuel or cladding damage.

Record the results of the estimated i

damage on the Attachment 11.

I CH-TI.ZZ-Oll(O) 4 Rev. O l

t

NOTE 5.2.3.8 Upon implementation of the Emergency Plan the Chemistry Engineer or his designee becomes the Chemistry Coordinator.

5.2.3.8 The Chemistry Coordinator shall sign and date Attachment 11 after verifying the computations.

5.2.4 Determination of Core Damage f rom Hydrogen Concentration in the Drywell Atmosphere.

5.2.4.1 Determine hydrogen concentration in the drywell atmosphere post-accident sample as % H2 IAW CH-EO.SH-004(0).

The drywell hydrogen monitor reading may be used as an alternative or for analysis comparision.

5.2.4.2 Using the curve for Mark I Containment on Attachment 13, determine the core damage (% metal-water reaction) for the reference plant.

NOTE 5.2.4.3 The % metal-water reaction for Hope Creek is assumed to be the same as the % metal-water reaction for the reference plant within the accuracy of this estimate.

5.2.4.3 Record the result on Attachment 11.

5.2.4.4 The Chemistry Coordinator shall sign and date Attachment 11 after verifying the computations.

5.2.5 Determination of Core Damage from the Drywell Radiation Level.

NOTE 5.2.5 This method for core damage estimation may be used in the event of substantial core damage.

CH-TI.ZZ-Oll(0) 5 Rev. O

5.2.5.1 Contact a control room operator and obtain a reading of the drywell radiation monitor from the Post Accident Monitoring Panel, (R), in R/hr.

5.2.5.2 Determine the elapsed time from plant shutdown to the drywell radiation monitor reading, (t), in hours.

5.2.5.3 Using Attachment 14, determine the fuel inventory release for the reference plant, (I)ref, in per cent.

NOTE 5.2.5.4 Formula and factors from Reference 7.3.

5.2.5.4 Determine the inventory release to the' drywell, (I), in % using the following formula:

(I)

(I)ref (1670)(

V

)(6/D)

=

P 237,450 where P = reactor power at time of incident, MWth i

V = total drywell free volume, 4

1.69ES ft3 D = distance of detector f rom reactor biological shield wall, ft.

5.2.5.5 Record the result on Attachment 11.

l' 5.2.5.6 The Chemistry Coordinator shall sign and date Attachment 11 after verifying the computations.

5.2.6 Identification of Release Source (Gap release or Meltdown Release) by Isotopic Ratios.

CH-TI.22-011(O) 6 Rev. 0

NOTE 5.6.2.2 Decay correct the concentrations to the sampling time, if necessary.

5.2.G.2 Calculate the isotopic ratios on attachment 15.

5.2.6.3 Compare the results of the isotopic ratio to the table on Attachment 15 to identify the release source.

5.2.7 Other Confirmatory Parameters 5.2.7.1 If isotopes of Sr, Ba, La and Ru are found in significant concentration, some degree of fuel melting may be inferred.

However, the extent of fuel melting cannot be determined based on*

the concentrations of these nuclides because of the lack of baseline data.

NOTE 5.2.7.2 Consideration should be given to the time the water level remained at that indication.

5.2.7.2 Contact the control room and obtain the Fuel Zone Range indication.

A reading of "0"

indicates no fuel damage,

"-50" indicates intermediate fuel damage and

"-100" indicates intermediate to severe fuel damage.

5.3 Calculations As required per Attachment 1 through 15.

5.4 Acceptance Criteria Not Applicable CH-TI.ZZ-011(0) 7 Rev. O

s 5.5 Reportino Results 5.5.1 The Chemistry Coordinator or his designeo shall report the results of Attachment 11 to the Emergency Director upon completion.

5.5.2 All results shall be re corded I AW CH-TI.ZZ-013(0).

6.0 ATTACHMENTS l

6.1, Decay Corrected Fission Product Concentrations 6.2,

Fission Products Inventory Correction Factors 6.3, Normalized Concentrations of the Fission Products 6.4, Xe-133 Concentration in the Drywell and Torus Gas to Cladding Damage i

6.5, Kr-85 Concentration in the Drywell and Torus Gas to Cladding Damage

?

6.6, Cs-137 Concentration in Primary Coolant l

to Cladding Damage 6.7, Cs-137 Concentration in the Primary Coolant and Torus to Cladding Damage 6.8,

I-131 Concentration in the Primary Coolant to Cladding Damage 6.9,

I-131 Concentration in the Primary Coolant and Torus to Cladding Damage l

6.10 Attachment 10, Example Sequence of Analysis 6.11 Attachment 11, Core Damage Estimates 6.12 Attachment 12, Hope Creek Plant Parameters / Ratios to Reference Plant 6.13 Attachment 13, Hydrogen Concentration in Drywell to Cladding Failure f

i l

l CH-TI.ZZ-011(0) 8 Rev. 0 l

f f

6.11 \\ r. ta chme n t 14, crywell Stonitor Dose Ra te to Fuel Inventory Airborne 6.15 Attachment 15, Identification of Release Source

7.0 REFERENCES

7.1 P&ID M-38-0, Post Accident Sampling System Sht. 1 of 2 and 2 of 2, Rev.

3.

7.2 CH-EO.SH-004(0), Post Accident Sample Analysis 7.3 NEDO-2215 82NEDO90, Procedure for the Determination of the Extent of Core Damage Under Accident Conditions 7.4 NEDC 30088, Responses to NRC Post Implementation Review l

Criteria for Post Accident Sampling System 7.5 USNRC Regulatory Guide 1.97, Instrumentation for Light Water Cooled Nuclear Power Plant to Assess Plant and Environs Conditions During and Following an Accident, 1980 t

7.6 N1-A41-38(1)-1 8010 1 (GEK-83344), Post Accident Sampling System, Operation Maintenance Instructions Volume 1, November 1981 t

7.7 INPO Good Practice EP-809, Guidance for Post-Accident Sampling, Preparedness, March 1985 7.8 Closing Documents 4

C D-3 8 5 Y, (HCGS FSAR Section 1.14.1.51)

CD-548X, (HCGS FSAR Section 9.3.2.3.2)

CD-443D, (VPN-PLP-09 Rev. O, Emergency Preparedness) 7.9 Hope Creek Generating Station Emergency Plan CH-TI.Z2-Oll(0) 9 (LAST PAGE)

Rev. 0

ATTACliMENT 1 DECAY CORRECTED FISSION PRODUCT CONCENTRATIONS l

Fission Product Concentrations Shutdown Time, To:

hrs Date Analysis Time, Ts hrs, Date De cay Time,

T= ( Ts-To ) :

hrs

24 '

day

Ci

= Fission Product Concentration of the Post Accident Sample.

i

= Decay Constant of the nuclide, day-1

Cit = Decay Corrected Fission Product Concentration.

Ait -

Cit

= Cie FISSION PRODUCT Ci li T

e AiT Cit I - 131 8.62 E-2 Cs - 137 6.29 E-5 Xe - 133 1.32 E-1 Kr - 85 1.77 E-4 concentrations expressed in uCi/g for liquids and uCi/cc for gases i

i s

Chemistry Activities Coordinator Date Time

-CH-TI.ZZ-Oll(0) 1 of 1 Re v.

0

._.- _ _.. _ _ ~_______

ATTACHMENT 2 FISSION PRODUCTS INVENTORY CORRECTION FACTORS Inventory Correction Factors (Fri Equation 1 Fri = Inventory in reference plant of Isotope i Inventory in operating plant of Isotope i

-1095)ki 3651 (1 - e

)

j[Pj(1 - e-i 3)e-i

]

where:

Pj

= Steady reactor power operated in period j, MWt

(+ 20% fluxuations in power are considered steady)

Tj

= Duration of operating period j, days o

Tj

= Time between the end of operating period j and time of the final reactor shutdown, days hk

= Decay correction factor of the isotope, days-1 i

3651 = Ave. Operation Power (in MWt) for the reference plant 1095 = Continuous Operation time (in Days) for the reference plant Using Equation 1, compute Fri for the fission products using i values provided in the following table.

A sample calculation is shown on Page 2 of Attachment 2.

Attach Calculation Sheets.

FISSION PRODUCT

)(i INVENTORY CORRECTION FACTOR, Fri I - 131 8.62 E-2 Cs - 137 6.29 E-5 Xe - 133 1.32 E-1 Kr - SS 1.77 E-4 CH-TI.ZZ-011(O) 1 of 2 Rev. O

.i...-

j l

ATTAOlMfNT 2 (cont'd)

FISSION PRODUCTS INWNIORY CDRRECTION FACIURS I

I Sampling Calculation of Fission Product Inventory Correction Factor, F Assuming a reactor has the following power operation history:

i Operation Operation Time Average Power o

Period Ihys Since Startup Tj (day)

Tj Pj (MWt) 1A 1 - 60 60 254 (= 314 - 60) 1000 1B 61-70 0

2A 71-270 200 44 (= 314 - 270) 2000 2B 271 - 300 0

3 301 - 314 14 0 (= 314 - 314) 3000 l

For I-131 ( A= 0.0862 day-1) 3651(1-e-0.0862x1095) i F (I-131) "

I 1000(1-e-0.0862x60)e-0,862x254 + 2000(1-e-0.0862x200) e-0.0862x44 + 3000(1-e-0.0862x14)e-0.0862x0 3651

=

= 1. 7 0 + 45 + 2103 For Cs-137 ( A= 6.29 x 10-5 day -1)

~

3651( l-e-6.29x10 x1095)

F (Cs-137)

I

-5

-5 1000(1-e-6. 29x10 x60)e-6.29x10 x254

-5

+ 2000(1-e-6.29x10 x200)e-6.29x10' x44 i

243.16

=7.77

=

-5

-5 i

+ 3000(1-e-6.29x10 x14 )e-6. 29x10 x0 3.74+24.93+2.64 cnanistry coordinator Date Time l

CH-TI.ZZ-011(0) 2 of 2 Rev. 0

ATTACIMNr 3 NOIM\\LIZED GEFETRATIONS OF 'IllE FISSION Pf0 DUCTS Ref 1.

Normalized Concentration of Primary Coolant Fission Products, Cwi

= Cit x Fri equation (A)

NorE 1.1, 1.2, 2.1, 2.2 The normalized concentrations shall be multiplied by Fw for liquid sanples and Pg for gas sanples if the system masses or volumes are different from those used in -9.

Calculations for f and a

F values are found on Attachment 12.

w 1.1 Normalized Concentration of I-131 Cit = uCi/g (From Attach. 1)

Fri = (Frm Attach. 2)

Substituting in equation (A) for I-131, Ref C

=

x xF (if necessary) w wi uCi/g

=

1.2 Normalized Concentratien of Cs-137 Cit =

uCi/g (From Attach. 1)

Fri =

(Frcm Attach. 2)

Substituting in equation (A) for Cs-137 Ref y

Cwi =

x xF (if necessary) y uCi/g

=

Ref 2.

NorTnalized Concentration of Containment Atmosphere Fission Products, Cn;

= Cit x Fri equation (B)

CH-TI.22-Oll(Q) 1 of 2 Rev. O

.5 s,

' ATrAOIMENE 3 (Cont'd)

NOIM\\LIZED CI)NGNIRATICN OF 'IllE FISSION Pf0 DUCTS 2.1 Nonnalized Concebtradon of Xe-133 C'it =

uCi,5$(FromAttach,1)

Fri =

(From Attach. 2)

Substituting for Xe-133, in ediation (B).

,3 Pef

Cgi-l=

x Fn (if necessary) uCi/cc

=

2.2 Normalized Concentration of Kr-85 Cit =

uCi/cc (Frcm Attach.1)

FI i _ =,, t (Fram Attach. 2)

Substituting for Kr-85, in equation (B) aee Cgi =

x-F (if necessary)'

g

=\\

uCi/cc

. \\

Record the calculated normalized concentrations in.the box-belcw.and use the

-information to compute core damage using Att:achments 4-9.

5 1

I-131 uCi/g Cs-137

.uCi/g uCi/cc Xe-133 Kr-85 uCi/cc Chemistry Activities Coordinator Date Time A

/mk4PJLSL92cA11/U60 W

N9hz@

Cli-!. : Z - : '.

ATTACHMENT 4 Xe-133 CONCENTRATION IN THE DRYWELL AND 'IORUS GAS 'IO CLADDING DAMAGE 100000--

i i

i Fuel Meltdown Upper Release Limit Best Estimate

/

10000

/ 7

/

//

/

//

/-

'/

1000

/f sj /

// /

ll

/

,7

/

100

/

c

.2

/

/ C1 adding Failure b

/

N c

/

Upper Release Limit

/

Best Estimate j

j

/

Lower Release Limit A

/

7

/

1 2

,/

/

1.0

/

t O.

0.1 1.0 10 100

" Cladding Failure

I i-I I

1.0 10 100 l-

% Fuel Meltdown

= 'i CH-TI.ZZ-Oll(O) 1 of 1 Rev. O

Zh-::.fl J'L,

ATTACHMENT 5 Kr-85 CONCENTRATION IN THE DRYWELL AND TORUS GAS TO CLADDING DAMAGE 1000 i

i i

Fuel Meltdown Upper Release Limit Best Estimate Lower Release Limit 100 -

/

///

/

/.

[/

/

10 -

l

/

// /

//, /

1.0-

/

l c2

/

l

/

{

/

Cladding Failure g

0.1

/

Upper Release Limit

/

[

/

Best Estimate 7

T

/

Lower Release Limit b

/

l

/

~

001

/

I 0001 O.1 1.0 10 100

Cladding Failure,

l 1.0 10 100 l-

% Fuel Meltdown

=l l

CH-TI.22-Oll(0) 1 of 1 Rev. O

(

~

CH-!I.22-01 ATTACHMENT 6 Cs-137 CONCENTRATION IN PRIMARY COOLANT TO CLADDING DAMAGE

100000, Fuel Meltdown Upper Release Limit j

Best Estimate

/

j Lower Release Limit

/

10000

/

/-

/

,I /

/

1000

/

,/ /

n b

/

,/

/

/ ll

/

co

/

l

/

E 100

/

b l

/

5

/

E l

/

r

/

l d

Cladding Failure 5

10 Upper Release Limit

[

Best Estimate Lower Release Limit 1.0

/

,/

/

0.1 O1 1.0 10 100 l

% Cladding Failure

=

1.0 10 100

% Fuel Mel tdown l

1

[

Attgegegt6 Re v.

0 CH-TI.ZZ-Oll(Q)

..:i- ::. _.-

ATTACHMENT 7 Cs-137 CONCENTRATION IN THE PRIMARY COOLANT AND TORUS TO CLADDING DAMAGE 10000 Fuel Meltdown Upper Release Limit

/>

Best Estimate wer Release Umi

/

A 1000 -

/

/ /.

/

/ l/

100 -

/

l 5

,/

/

/-

l

/

8

/

B

/

2

/

7' 10 -

0

/

b

/

~

/

m h

/

3

/

Cladding Failure y

1.0/

Upper Release Limit

/

Best Estimate Lower Release limit

/

0.1 f

/

0.01 O.1 1.0 10 100

% Cladding Failure

={

=

1.0 10 100

% Fuel Meltdown

.CH-TI.ZZ-Oll(O) 1 of-1 Rev. 0

JH-::.22-0.1 ATTACHMENT 8 I-131 CONCENTRATION IN 'IHE PRIMARY COOLANT 'IO CLADDING DAMAGE

't 1000000 Fuel Meltdown Upper Release Limit S :t E: tim:tc p

100000

/ 2

/

/ /

/

/ /

10000 g

7

/

/ l/,/

k b

/

/ /

/

c l

/

j 1000 7

/

c8

/

7

/

l

]

100 -/

/

p

/

Cladding Failure f

Upper Release Limit

/

Best Estimate 10

/

Lower Release Limit

/

^

/

I 1.0 O.1 1.0 10 100 l-

% Cladding Failure

l 1,. 0 10 100

% Fuel Meltdown g

CH-TI.zz-Oll(0)

CH-!

.'J-i ATTACHMENT 9 I-131 CONCENTRATION IN THE PRIMARY COOLANT AND TORUS TO CLADDING DAMAGE 100000 g

Fuel Meltdown Upper Release Limit Best Estimate

/

7 Lower Release Limit

/

10000

/ \\

l/l

/ /

/ / I

/

1000 p,l lj l /

[m 5

/

~

5

//

l

/

B

/

b

/

g 100 8

/

u A

/

2

/

.5 j

e 3

10 j'

7

! addino Failure Cl

/

Upper Release Limit

/

Best Estimate Lower Release Limit j

/

l 0.1 O.1 1.0 10 100

{

% Cladding Failure j

1.0 10 100 l=

% Fuel Meltdown

{

CH-TI.ZZ-Oll(Q) 1 of 1 Rev. 0

C ri - T I. ? Z ;...

ATTACHMENT 10 EXAMPLE SEQUENCE OF ANALYSIS H015GL OPERATIOR Hydrogen Containment Water I"

- MINOR CLAD DAMAG Analysts M Radiation - M Level (Confire)

(Confire)

(Conffm)

}

3 0, r.ine i Cor. 0.-,e ;

Optimum Estimete' n

l,'

Sample From PASS Point Analysis For Hydrogen Containment Water yg Ba. Sr. La. Au My Level.

Analysis M,

Radiation (Conftre)

(ConfIre)

(Conffrm)

-i MM0R O.A0 DAMAE Detereinstion FIEL OVEllHEAT g

Of Fisston Product Retfos FUEL M LT En CLAD DAMAE POSSIBLE FUEL OVEINEAT NO CORE ELT_

10 Rev. 0 A

qggegt CH-TI.22-011(0)

ATTACHMENT 11 CORE DAMAGE ESTIMATES 1.

Core Damage Estimates based in the Fission Products Concentrations in the post accident samples Record results obtained based on concentrations reported in and using Attachments 4-9 in the following table.

FISSION PRODUCT

% CLADDING FAILURE

% FUEL MELTDOWN I-131 Cs-137 Xe-133 Kr-85 2.

Core damage estimate based on the Hydrogen Concentration in the containment Record results obtained per Step 5.2.4.2

% MW reaction

(% Cladding Failure) 3.

Core damage estimate based on radiation level in the drywell caused by airborne fuel inventory of fission products per Step 5.2.5.4

% Core Damage

(% Inventory airborne) 4.

Summary Final Estimation Chemistry Coordinator Date Time 1 CH-TI.ZZ-Oll(0) 1 of 1 Re v. O

0;_

r,, 3 3 a

ATTACHMENT 12 HOPE CREEK PLANT PARAMETERS /Fw AND Pg CORRECTION FACTORS HOPE CREEK PLANT PARAMETERS:

Rated Reactor Power 3293 MWt Number of Fuel Bundles 764 Steam Flow at Full Power 1.78E3 kg/s (14.16E6 lbs/h)

Reactor Coolant Mass At Power 2.93E5 kg (6.44E5 lbs)

-Hot Standby 3.03E5 kg (6.69ES lbs)

. Cold Shutdown 4.09ES kg (9.02E5 lbs)

Torus Water Volume (MASS) 3.34E6 kg (1.18E5 ft3)

Free Volume 3.78E6 liters (1.34E5 ft3)

Drywell Free Volume 4.79E6 liters (1.69E5 ft3)

NOTE If the isotopic concentrations of the drywell and torus are perceived to be equal, Fg and Fw may be considered as equal to 1.

Fw AND Fg CORRECTION FACTORS:

Containment Gas Volume Correction Factor (Fq)

Hope Creek torus and/or drywell gas volume, (

cc)

Hope Creek Torus & Drywell Gas Volume (8.57E9cc)

Primary Coolant Mass Correction Factor (Fw)

System Coolant Mass (

g)

Fw =

=

Primary Coolant Mass (3.03E8g) reactor coolant or (3.64E9g) torus & reactor coolant 2 C H-TI. Z Z-011( O) 1 of 1 Rev. 0

.: H - : :. Z Z '; ; l 1

ATTACHMENT 13 HYDROGEN CONCENTRATION IN DRYWELL 10 CLADDING FAILURE a

04 so se j

E 4

I aa E

e

=

wlm 1#

=

s.

u

=

24

=

20 I

i.

12 =

a-4 -

8 i

i a

a n

0 I

M M

50 80 70 80 go 100

% METAL-WATER REACTION NUREQ4737 4 3 CH-TI.ZZ-011(0) 1 og 1 Rev. 0

E

H-::.2?-);

ATTACHMENT 14 DRYWELL MONITOR DOSE RATE TO FUEL INVENTORY AIRBORNE Percent of Fuel 2nventory Airborne in the Centaiansent Y l 2004 Puel Inventory = 1004 Noble Gases g,1

+ 234 2edine

+ 14 particulates

) M)

L loot Q:

eggf' a;

1 lot 4 be 5 L

3, e

?*

g 1

0.1 4

Y 0.01 I b'

O.0014 k' 7 10" ) 3 b iDWit ) ) i b HWig ) ) i hh'r0Fif i ) i h bWto' nme After Skuldewrn (Hrs)

% Fuel Inventory Approximate source and Damage Estimate Released 100.

2004 T2D-14044, 2004 fuel damage, potential core melt.

50.

50% T2D amble gases, sur seeres.

10.

104 TID, 2004 NRC gay activity, total elad failure, partial sore aseovered.

3.

34 TID, 2006 Whss-1400 gay activity, major elad fallare.

1.

14 T2D,104 MC gay, ans.106 eled fallare.

.1

.1% TID, It MC gep, Il eled fallere, leeal heating of 5-10 fuel assemblies.

.01

. tit feel. TID,.It' WRC gap, eled failure of 3/4 l

alment (34 reds).

l I

10-1

.014 nac gay, elad failure of a few reds.

10-4 1906 esolant release with sylking.

de10*d 1000 emelant lassatory release.

13-0 ppper range et assual airherme amble ges aeolvity la samtalaneet.

l 4 CH-TI.ZZ-Oll(0)

I f1 Rev. O i

L k

4

l-CH-TE.22-Olif

.e ATTACHMENT 15 IDENTIFICATION OF RELEASE SOURCE MELTDOWN RELEASE GAP RELEASE ACTIVITY RATIO

IN ACTIVITY RATIO

IN ISOTOPE HALF-LIFE CORE INVENTORY FUEL GAP Kr-87 76 m

0.233 0.0234 Kr-88 2.84h 0.33 0.0495 Kr-85m 4.48h 0.122 0.023 Xe-133 5.25d 1.0*

1.0*

I-134 52.6 m 2.3 0.155 I-132 2.28h 1.46 0.127 I-135 6.59h 1.97 0.364 I-133 20.8 h 2.09 0.685 I-131 8.04d 1.0*

1.0*

SHOW CALCULATIONS:

Kr-87 conc. (

)

I-134 conc. (

)

=

Xe-133 conc. (

)

I-131 conc. (

)

Kr-88 conc. (

)

I-132 conc. (

)

=

Xe-133 conc. (

)

I-131 conc. (

)

Kr-85m conc. (

)

I-135 conc. (

)

=

Xe-133 conc. (

)

I-131 conc. (

)

I-133 conc. (

)

=

I-131 conc. (

)

Ratio = noble gas isotope concentration for noble gases Xe-133 concentration

= Iodine isotope concentration for iodines I-131 concentration RELEASE SOURCE

/

Performed by Date Chemistry Activities Date Coordinator 5 CH-TI.ZZ-011(0) 1 of 1 Rev. O

ML20127J356

Text

7.0 REFERENCES

7.0 REFERENCES

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