ML20154J992

Forwards Portion of Design Document for Emergency Dose Projection Sys,In Response to Telephone Request for Emergency Source Term Calculational Methodology

ML20154J992
Person / Time
Site:	Columbia
Issue date:	02/26/1986
From:	Sorensen G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Sherman C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
NUDOCS 8603110144
Download: ML20154J992 (6)

Text

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O Washington Public Power Supply System P.O. Box 968 3000 GeorgeWashingtonWay Richtand, Washington 99352 (509)372-5000 Docket Number 50-397

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February 26, 1986 g

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q g*1 Mr. Conrad Sherman U.S. Nuclear Regulatory Consnission 1450 Maria Lane Suite 210 Walnut Creek, CA 94596

Subject:

DESIGNDOCUMENTFOREMERuENCYDOSEPROJECTIONSYSTEM(EDPS)

In response to your telephone request of G. V. Oldfield for emergency source term calculational methodology, we are enclosing a portion of the design document for our Emergency Dose Projection System (EDPS).

It describes the methods used to calculate the reactor building source term based upon elevated release point (ERP) high range monitor read-ings.

The calculational method initially assumes a design basis analysis mixture of fission products in the reactor building (FSAR Table 15.6-13).

This mixture may be modified if grab sample analysis results are available. The EDPS program corrects the mixture for decay from the time of reactor shutdown then calculates the ERP monitor response to that mixture using nuclide specific calibration factors. The actual ERP monitor reading is divided by the calculated response to generate a normalization factor. The design basis ccncentrations are multiplied by the normalization factor to calculate actual release concentrations.

Release concentrations are multiplied by thri flow rate to determine release rate.

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o Mr. Conrad Sherman Page 2 February 26, 1986 Docket Number 50-397 Similar calculational methods are used to calculate source terms from other monitor readings (intennediate and low level) in the reactor, turbine generator, and radwaste buildings..If you have any questions concerning these methods, please feel free to contact Mr. Oldfield.

Sincerely, O

G. C. Sorensen, Manager Regulatory Programs GV0/pm Enclosure

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DESIGN OF SUBROUTINE PREDCAL-IN EDPS 1.

Tables are defined the same as in the CALC phantom in EDPS. They are:

,%.++rr ;. c.n w,i m.v.,4 (ptA5 ~~.,ye ; e :,.;;)

.DesignBase' Analysis (Ci) a.

b..

Half-life (min)

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404 7 3

c.

Elevated Release Point Location Monitor Calibration Factor (R/Hr per Ci/m )

3 d.

Wale Body Dose Factor (R/Hr per Ci/m )

3 e.

Thyroid Dose Factor (R/Hr per C1/m )

' 2. n--If both-MeLand Effluent aregt in predictive mode, write an er,ror message

'and~ return.

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3.

Zero out predictive effluent arrays.

4.

Read in predictive effluent arrays that were stored with user's responses.

5.

If source term for all three buildings < 0.001, write an error message l

and return.

6.

Get times in minutes as entered by the user.

a.

Elapsed time is decay time (time from reactor shutdown until release).

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b.

Delay time is the duration of the ' release.

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7.

Replace the values for the iodines in the DBA table with grab sample values.

If there were predictive grab samples read in, use those; otherwise, use manual ones, i

l 8.

Calculate D(I) = Decayed activity in primary containment of Ith nuclide.

l Using the formula:

D(I) = A(1) e-0.693M/T(I) l Where:

A(I) = Initial airborne activity of nuclide I (from Design Base Analysis l

Table) l M=

Elapsed time (min.)

t-D ' *" m "*

" W/N h" *" " # #

l T(!) = Half life of Nuclide I in minutes YI" I'/""

N If /T(!) > 50, Set D(I) = 0 i

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e sa y 9.1 CalculateC(I)=concentrationofIthnuclidein, containment Using the' formula:

C(I) = D(I) x 9.10 x 10-6 (Ci,3) j 10.

Calculate contribution to measured exposure rate in elevated release duct from nuclide I at concentration C(I).

In case of iodines, concentration L

in duct will be reduced by SGTS.

.C(I) duct = C(I) containment x (1-0.015)

Where S = % removed by SGTS (either 0 or 99.7) l

..11.. Calculate E(I) = exposure rate calculated at the ERP Monitor resulting r ;from-Nuclide I in R/Hr Using the formula:

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E(I) = C(I) duct x F(I) l.

Where F(I) is from the Calibration Factor table.

12. The total exposure rate calculated at the ERP Monitor (x) would be given by:

l 18 x=I E(I)

!=1

13. Make determination whether to use high or intermediate range monitor for reactor building.

Use intermediate range if its readingd100 and high range reading 5.5.

Otherwise, use high range.

If high range (ion chamber) used.,.

a.

The actual exposure rate measured by the ERP Monitor (R) is divided by the calculated exposure rate to determine a normalization factor (N),

g,R/X The source term for each nuclide Q(I), is calculated by multiplying

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C(I) duct by the flow rate, G, and the normalization factor, N.

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Q(I)Ci/sec. = C(I) Ci/,3 xg

/ min. x N x 4.72 x 10~4 3

m min i

ft.3sec l

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DESIGN OF CALC PHANTOM IN EDPS 1.

The following tables are defined:

ERP Loc. Mon.

Whole Body Thyroid Design Base Hal f-Li fe Calib. Fact.

Dose Factor Dose Factor 3

3 3

Nuclide Analysis (01)

(Min)

(R/Hr per Ci/M ),

(R/Hr per Ci/m )

(R/Hr per Ci/m )

1-131 2.2 E+7 1.7.1 E+4 3.35 242 3.42 E46 I-132 3.3 EF7 137.0 24.4 1.43 E+3 4.08 E+4 I-133 4.9 E+7 J1.25 F+3 4.0 383 8.11 E+5

. 1-134 5.6 E+7 52.5 21.9 1.65 E+3 1.07 E+4 I-135

- 4. 4 E+7_.

397.

14.9 1.04 E+3

• - 1.67 E+5 Kr-83M 1.4 E+7 110.

9.94E-9 8.62 E-3 0

Kr-85M 4.5 E+7 269.0 1.23 133 0

Kr-85 1.4 E+6 5.63 E+6 0.0254 1.84 0

Kr-87 8.0 E+7 76.3 11.8 675 0

Kr-88 1.1 E+8 172 13.3 1.68 E+3 0

Kr-89 1.1 E+8 3.18 16.7 1.89 E+3 0

Xe-131M 9.0 E+5 1.69 E+4

0. 0218,3,4 -< 10. 4 0

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Xe-133M 4.8 E+6 3.16 E+3

((0.0298]!' ' 28.6 0

Xe-133 1.9 E+8 7.55 E+3 0.0766 33.5 0

Xe-135M 5.1 E+7 15.7 3.55 356 0

Xe-135 1.9 E+8 546.

2.13 206 0

Xe-137 1.5 E+8 3.82 1.86 162 0

Xe-138 1.6 E+8 14.2 12.5 1.01 E+3 0

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TABLE 15.6-13 i

LOSS-OF-COOLANT ACCIDENT (DES 1CN BASIS ANALYSI$)

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l ACTIVITY AIRBORNE IN PRIMARY CONTAIt#ENT (CURIES)

ISOTOPE I MIN 30 MIN Im 2m 4m 8 HR 12 m i DAY 4 DAY 30 DAY l-138 2.2E 07 2.2E 07 2.2E 07 2.2E 07 2.lE 07 2.lE 07 2.lE 07 2.0E 07

' l.5E 07 f.6E 06 l-132 13E 07 2.8E 07 2.4E 07 f.8E 07 9.8E 06 2.9E 06 8.6E 05 2.3E 04 7.1E-06 0.

1-133 4.9E 07 4.8E 07 4.7E 07 4.5E 07 4.3E 07 1 7E 07 3.3E 07 2.2E 07 2.0E 06 1.9E-03 I-134 5.6E 07 A LE 07 2.6E 07 1.2E 07 2.4E 06 1.0E 05 4.2E 03 1 2E-01 0.

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l-135 4.4E 07 4.2E 07 4.0E 07 3.6E 07 2.9E 07 1.9E 07 1.2E 07 3.5E 06 f.8E 03 0.

j TOTAL 2.0E 08 8.8E 08 f.6E 08 1.3E 08 1.1E 08 8.0E 07 6.7E 07 4.5E 07 1.7E 07 1.6E 06 Kr-83e 1.4E 07 f.2E 07 9.9E 06 6.8E 06 1 2E 06 7.2E 05 1.6E 05 1.8E 03 3.4E-09 0.

Kr-85m 4.5E 07 4.2E 07 3.8E 07 3.3E 07 2.4E 07 1.3E 07 7.0E 06 f.lE 06 1.5E 01 0.

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Kr-85 1.4E 06 f.4E 06 1.4E 06 1.4E 06 f.4E 06 f.4E 06 f.4E 06 f.4E 06 1.4E 06 f.4E 06 w

Kr-87 8.0E 07 6.1E 07 4.7E 07 2.7E 07 9.0E 06 1.0E 06 1.lE 05 f.6E 02 0.

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l Kr-88 1.lE 08 9.8E 07 8.6E 07 6.7E 07 4.lE 07 1.5E 07 5.7E 06 2.9E 05 5.IE-05 0.

Kr-89 f.lE 08 f.9E 05 2.6E 02 4.9E-04 0.

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Xe-13 tm 9.0E 05 9.0E 05 9.0E 05 8.9E 05 8.9E 05 8.8E 05 8.7E 05 8.5E 05 7.lE 05 1.6E 05 1

Xe-133m 4.8E 06 4.8E 06 4.7E 06 4.7E 06 4.5E 06 4.3E 06 4.lE 06 3.5E 06 f.4E 06 4.3E 02 Xe-133 8.9E 08 1.9E 08 1.9E 08 f.9E 08 8.9E 08 f.9E 08 8.8E 08 f.7E 08 1.IE 08 3.8E 06 Xe-835m 5.lE 07 1.4E 07 3.6E 06 2.3E 05 f.0E 03 f.9E-02 3.7E-07 0.

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O.

Xe-135 1.9E OJ I.8E 08 f.7E 08 f.6E 08 f.4E 08 1.0E 06 7.5E 07 3.0E 07 f.3E 05 O.

Xe-137 1.5E 06 7.8E 05 3.5E 03 6.8E-02 2.6E-Il 0.

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O.

O.

O.

4 Xe-138 f.6E 08 A SE 07 8.8E 06 4.7E 05 1.3E 03 1.lE-02 8.8E-08 0.

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C H *n MZ TOTAL NG l.lE 09 6.5E 08 5.7E 08 4.9E 08 4.IE 08 1 2E 08 2.8E 08 2.IE 08 f.2E 08 5.4E 06 e $,

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