ML20140D573

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Rev 1 to Radiological Assessment of Spent Fuel Shipping Cask Drop in Cy Sfp
ML20140D573
Person / Time
Site: Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date: 03/13/1997
From: Landry L, Miller D
HOLTEC INTERNATIONAL
To:
Shared Package
ML20140D536 List:
References
XX-XXX-60RA-R01, XX-XXX-60RA-R1, NUDOCS 9706100417
Download: ML20140D573 (32)
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CALCULATION TITLE PAGE 28 Number of cale. pages

'. Number of attachment pages 4 Total number of pages 32 Radiological Assesstnent of a' Spent Fuel Shipping Cask Drop in the CY Spent Fuel Fool TITLE Note: Rev. I also includes the Fuel XX-IIX-60RA 1 Handling Accident CALCULATION # REV.# Vendor Calc #  !

1 System StructureSpent Fuel Bldg. Component Executive Summary l

.= I This calculation is a revision to the original (rev. 0), tne changoa in assumptions are given on page 4 under the section - Purpose. .

The effective date of this calculation is the date of approval because j

radioactive decay was considered to occur from July 22, 1996 to February 28, 1997 (221 days).

The radiological consequences of a Cask Drop Accident in the CY Spent CY Spent Fuel Pool are l 30% of 10CFR100 limits)given on page 4 and are acceptable (eg. less than Does this calculation:

1. Support a DCR, MMOD, an mdependent review method for a DCR ,or confirm test Yes ONoD results for an installed DCR7 If yes, indicate the DCR, MMOD number and/or Test Procedure number.
2. Support mdependent analysis? If yes, indicate the procedure, work control or other Yes ONo @

reference it supports.

3. Revise, supersede, or void enstmg calculanons? If yes, indicate the calculation number Yes ONo O and revisions. XX-XXX-60RA. Rev. 0
4. Involve QA or QA-related systems, comporants or structures? Yes ONo O
5. Impact the Unit licensmg basis, including technical specificanons, FSAR, procedures or Yes ONo O licensing comnutments? If yes, identify appropriate change documents.

Approvais (Pnnt/ Signature) , ,A T M er (Pnnt/ Signature) Donal A Wdkr AM M Date: 03/05/91 Independent Reviewer (Pnnt/ Signature) L 4,n La ndry (My De Date: G/e/$1 f

Supervisor (Pnnt/ signature) em A (bEJ h4 /MAh.// Date: i //2 / ? 7 e

213 NUC DCM FORM 5-! A P PDR Rev.04 l f

Cat. 1 Calc. # l XX-XXX-60RA, Rev. 1 )

page 2 of 28 l Table of Contents 1

I) Purpose of Calculation 4 l 1

l l II) Summary of Results 4 1 1

III) References 5 ,

1 IV) Assumptions 7 l l

V) Method of Calculation 8 VI) Body of Calculation 13 l VII) Reviewer's Comments and Resolution 23 l l

VIII) Comments by CY Staff 24 I IX) Comments by Independent Reviewer 26 X) Attachments 28 ,

1 A) Attachment A -I-131 Dose Contribution Al l l

B) CTP Data Base Form El l l

C) Telecon C1 l l

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Cat. 1 Calc. #

XX-XXX-60RA,Rev. 1 r

page 3 of 28 List of Tables Table 1 - Assumptions 20 Table 2 - Calculation of Fuel Decay Time 22 l

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Cat. 1 Calc. #

xx-xxx-60 RA, Rev. 1 Page 4 of 28 I)

Purpose:

The purpose of this calculation is to reanalyze the fuel handling and cask drop accidents in the CY spent fuel pool for changes in the following assumptions:

1) Number of assemblies damaced: The previous analysis, Category 1 Calc. # xx-xxx-60 RA, Rev. O, assumed 400 assemblies would be ruptured by the cask. The decay times were variable according to the actual decay histories of the assemblies in the spent fuel pool. This analysis assumes one full core of assemblies are damaged (this is conservative because ref. 11 shows that a maximum of 132 assemblies could actually be damaged) .

This analysis also decayed these assemblies from July 22, 1996 (ie, date reactor was shutdown) to February 28, 1997 (The approval date which will be after February 28, 1997 is the date when this calculation will be effective).

2) Thyroid dose calculated from I-129 and skin from Kr-81,:

The previous calculation (referenced above) did not calculate the thyroid dose from I-129 nor the skin dose from Kr-85. This calculation does calculate those doses.

II) Summary of Results:

The results of this calculation are as follows:

One Assembly The doses assuming one assembly is damaged are:

Exclusion Area Boundary: whole body dose = 2.60E-03 (rem)

I thyroid dose = 2.28E-04 (rem)

[ assuming DF=100) thyroid dose = 2.28E-02 (rem)

[ assuming DF= 1) skin dose = 2.16E-01 (rem)

Low Population Zone: whole body dose = 2.01E-04 (rem) i thyroid dose = 1.77E-05 (rem)

(assuming DF=100) thyroid dose = 1.77E-03 (rem)

[ assuming DF= 1) skin dos 1.67E-02 (rem) c >

Cat, 1 Calc- #

XX-XXX-60RA, Rev. 1 page 5 of 28 Cask Droo (157 assemblies)

The doses assuming 157 assemblies are damaged are:

Exclusion Area Boundary: whole body dose = 4.08E-01 (rem) thyroid dose = 3.58E-02 (rem)

[ assuming DF=100]

thyroid dose = 3.58E+00 (rem)

[ assuming DF= 1]

skin dose = 3.4E+01 (rem)

Low Population Zone: whole body dose = 3.16E-02 (rem) thyroid dose = 2.78E-03 (rem)

[ assuming DF=100]

. thyroid dose = 2.78E-01 (rem)  ;

[ assuming DF= 1]  :

skin dose = 2.63E+00 (rem)

The doses summarized above are under the Standard Review Plan 15.7.4 limits of 75 rem thyroid and 6 rem whole body ,

and are therefore acceptable.

Note: Since the half life of I-129 is 10' years and the half life of Kr-85 is 10 years, the calculated doses will not change significantly over the subsequent years'. The importance of the 221 days decay was to ensure all average half life nuclides (eg. I-131 7 days) have decayed to in-significant levels. Suybsequent decay is not important.

III) References

1) Memo, Edward Mullarkey to Raymond Crandall,

Subject:

The Connecticut Yankee Spent Fuel Pool Rerack Program, i Cask Drop Analysis, DECY-96-0229, April 12, 1996 l

2) Connecticut Yankee Atomic Power Company Docket No.

l SD-213 Haddam' Neck Plant Facility Operating License, l license No. DPR-61, Date of Issuance : December 27, 1974.

l

3) Connecticut Yankee Updated Final Safety Analysis Report, October 23, 1996.
4) Safety Guide 25, Assumptions Used For Evaluating The Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for j ' Boiling and Presuurized Water Reactors, US Atomic mJ

. - . - . . = - - . - . _ . - _ .- - _ - . - - - . .-._ - .- - _ .

Cat. 1 Calc. #

XX-XXX-60RA, Rev. 1 page 6 of 28 Energy Commission, March 23, 1972.

5) TACT III Atmospheric Transport Code System, Oak Ridge National Laboratory, CCC-447.
6) Bucher, Sean W.,

Connecticut Yankee Cycle 19 Isotopics:

End of Cycle - 07/22/96, Cat. 1 Calc. No. CY19 TOTE-01502-FY, Rev. O, 08/30/96.

7) Santovasi, J. A.,

Chi /Q's at Site Boundaries and LPZ l (4345 m) for Release from Outside Face of Containment ,

l sna Rev. O, Convainment Vent Stack, Cat. 1 Calc. # XXXXX-30-PS, September 24, 1981.

l l 8) Telecom between Jay Lee, NRC and Donald Miller, NUSCo, l January 21, 1997.

l

9) Calculation of Annual Doses to Man From Routine Releases .

of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Regulatory.

l Guide 1.109, Rev. 1, US Nuclear Regulatory Commission,

! October, 1977.

10) Memo, J.J. Parillo to M.S. Kai,

Subject:

Reload and Safety Analysis Checklist (RSAC) for Connecticut i

Yankee, Cycle 19, Revision 0, NE-94-F-311, June 24, 1994.

11) Holtec International Report, HI-951308, " Analysis of Postulated Cask Drops Over The Connecticut Yankee Spent Fuel Pool".
12) Memorandum, From: J. DiStefano, To: J. Bourassa, l

Subject:

"QA Review of CY Spent Fuel Cask Drop Accident Analysis", File # CYJDS97.WPD, Yankee Atomic - Bolton, February 7, 1997.

13) Memo, From: Stephen J. Milioti, To: Don Miller,

Subject:

" Comments on Radiological Assessment of a Spent. Fuel Shipping Cask Drop in the CY Spent Fuel Pool", DECY 97-0537, February 11, 1997.

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Cat. 1 Calc. #

XX-XXX-60RA, Rev. 1 page 7 of 28 r

IV) Assumptions:

The fuel handling and cask drop accidents were assumed to occur after February 28, 1997 (ie. 221 days decay from July 22, 1996). The only gaseous isotopes that were assumed to be remaining were I-129 and Kr-85. The release was assumed to be an unfiltered ground level instantaneous release from the fuel building at 221 days after shutdown.

Atmospheric dispersion factors for a ground level containment elease were used as these were readily I available.

. l The largest ft 1 assembly burnup for the assemblies stored I l in the spent fuel pool 's 42,955.5 MND/Mt (ref. 6,page C2).

This corresponds to asset,51y number W 47. Isotopes Kr-85 and I-129 have long-half A f ves so- the activity in the assemblies will tend to bui..' dup over a long period of time (eg..these isotopes wil:. not be in secular equilibrium). The activity fcr these isotopes will be dependent on fuel assembly burnup. The TACT Library values for specific activity have variable burnups ranging in value from an average of 23,000 MWD /MT to a high of 40,000 MWD /MT (ref. 8). If we assume that Kr-85 and I-129 were

! based on the average burnup of 23,000 MWD /MT, we would I be conservative to multiply the specific activity of the ,

TACT library values by a factor of two. The TACT library i

! specific activity values for Kr-85 and I-129 were I

therefore multiplied by a factor of two.

J The assumptions used in this analysis are listed in Table 1 along with the basis of the assumption.

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Cat. 1 -Cal'c . # l XX-XXX-60RA, Rev. 1 page 8 of 28 1

l V) Method of Calculation:

1 Since there are only two isotopes of concern (ie. I-129 and 1 Kr-85 ), a hand calculation will be used to determine the l thyroid, whole body and skin doses. The equations used are:

A) Doses from One Assembly

1) Thyroid Dose from I-129 D;3. cx, = A, x DCF:.u, x BR x X/Qta,

)

D:H-trz" A p x DCF I-u, x BR x X/Qa ,

1 where: Drs.tx,= Thyroid dose at the exclusion area boundary (rem) from one assembly damaged D:3.u:= Thyroid dose at the Low Population Zone (rem)  ;

A, = I-129 activity that reaches the spent fuel pool surface from a damaged assembly (see below for definition of 5 9)

DCF -u, = Adult thyroid dose conversion 2

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factor for I-129 BR = breathing rate X/Qt, = Exclusion Boundary atmospheric i dispersion coefficient l Deny.tp = Thyroid dose at the Low l Population Zone I

A, = Pr x f, x 1/N, x 1/Dr xAsp z-us X tP x e' ' (3) where: A,= I-129 activity that reaches the spent fuel pool surface from a damaged assembly t

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I Cat. 1 Calc. #

XX-XXX-60RA, Rev. 1 page 9 of 28 P, = Radial peaking f actor for l

assembly f, = fraction gap of I-129 activity in the Na = total number of assemblies in the reactor core

- D,= spent fuel pool decontamination factor A sp er-u, -=

total specifice activity of I-129 that is in the reactor core Pt = reactor power level i

lz-u, =

decay constant for I-129 (1/hr) t = time (hrs) from shutdown to February 28, 1997. Table 2 calcu-lates the number of days decay starting.from shutdown on July 22, l 1996 (ref. 6, page 4) until Feb.

28, 1997 (arbitrarily selected).

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2) Whole Body Dose from Kr-85 Dw,.us = A, x DCFa.,3 x X/Qu, (4)

Dw, te, = A, x DCFa.,3 x X/Q,, t (5) l where: Dw,u,= whole body dose at the l Exclusion Area Boundary from I one assembly damaged (rem)

Ap = Kr-85 activity released from damaged assembly DCFu.,3 =

semi-infinite cloud whole body l dose conversion factor '

l (rem-m'3 ) / (Ci-sec') for Kr-85 l

x/Que =

Exclusion Area Boundary ground I level atmospheric dispersion ,

coefficient (sec/m'3) '

De :n - Whole body dose at the Low Population Zone o

Cat. 1 Calc. # -

XX-XXX-60RA, Rev. 1 page 10 of 28 x/Q tp = Low Population Zone ground level  ;

atmospheric dispersion ,

i coefficient (sec/m'3) I i

l A, = P, x f, x 1/Nx x A,x Pt xe -A* (6) where: A, x,. 3 = Kr-85 activity which reaches t

the spent fuel pool surface.

l All rods in one assembly are assumed to be damaged and Kr-85 gap activity is assumed to be released in pool.  !

f, = fraction of activity in fuel ,

rod gap A, = specific core activity of Kr-85

, (Ci/Mwt)

Pt = reactor power level (Mwt)

A = decay constant for Kr-85 t = time from shutdown (July 22,  !

1996) to February 28, 1997 l

! 3) Skin Dose From Kr-85 Dsr un = Dv,un x DCFsk n.as/DCFa. 3 (7)

I Dsx tra = Dws t,2 x DCFsr n es/DCFa. 3 (8) where: Dsr us= Skin dose at the Exclusion Area )

Boundary form one assembly damaged i (rem)

DCFx 3 n.3= semi-infinite cloud skin dose conversion factor for Kr-85 '

(rem-m^3)/(Ci-sec). 1 Dsr ter= Skin dose Pt the Low Population f Zone from one assembly damaged (rem) .

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l all other parameters are as defined previously l

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l Cat. 1 Calc. #

XX-XXX-60RA, Rev. 1 page 11 of 28 B) Doses From Cask Drop The doses from a. cask drop accident can be easily calculated from the doses calculated for a fuel assembly by using the following equations:

Note: These results are conservative as this method assumes all assemblies in the core experience a peaking factor of 1.7.

1) Thyroid Dose From I-129 Drx u3 c.,x = Drs ,,3 x Rx (9)

Dru t, c . = Drn tp xR3 (10) 1 where: Dru na essa = Thyroid dose at Exclusion area boundary from a cask  ;

drop accident in the spent fuel pool (rem)

Drs zu = Thyroid dose calculated for ,

one damaged assembly using i equation (1) (rem)

Rx = number of assemblies I ruptured by cask (conser- l vatively taken to be the  !

entire number of assemblies in the last core because l the calculated value of 132 damaged assemblies was very close to the total number of assemblies in a core)

2) Whole Body Dose from Kr-85 Dwn unc.s= = Dw un x Rx (11)

Dw, t, e.,, = Dwn t, x Rx (12) i where: Dwn trz c... = Whole Body dose (rem) at I

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Cat. 1 Calc. # '

l XX-XXX-60RA, Rev. 1 page 12 of 28 t

l Exclusion Area Boundary from a cask drop accident in the spent fuel pool i Dws un cesk = Whole Body dose (rem) at the Exclusion Area Boundary }

for one damaged fuel l assembly using equation (4) l Rx =

number of assemblies ruptured by cask (conser-vatively taken to be the entire number of assemblies in the last core because the calculated value of 132 damaged assemblies was very ,

close to the total number 1

of-assemblies in a core) i i

3) Skin Dose From Kr-85 I 1

Dsx cAsr cAs = Dwa cAsx EAs x DCFsx xr-ss/DCFra. 3 (13)

Dsx cAsx trt " % cAsx trz x DCFxga.3/DCFrm.3 3 (14) l l

where: Dsx cask tAs = Skin dose at the Exclusion Area Boundary from a cask drop in the spent fuel l pool (rem)

Dsr cur tra = Skin dose at the Low Popslation Zone from a cask drop in the spent ,

fuel pool (rem) l All other parameters as previously defined.

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Cat. 1 -Calc . #

XX-XXX-60RA, Rev. 1 page 13 of 28 VI) Body of Calculation A) Doses from One Assembly i i

1) Thyroid Dose from I-129 From equation (3) , we have:

Ap = Pg x f, x 1/N, x 1/Dr xAsp 2-12, xPt x e-At where: Pr = 1. 7 f,= 0.3  ;

NA 157 assemblies  !

Dr = 10 0 i f

4 Asr 2-12,=

9.26E-04 Ci/Mwt x 2 = 1.85E-03 Ci/Mwt P 2 , = 18 25 (Mwt)

{

l = 5.04E-12/hr ,

t = 221 days Ap =

1.7 x 0.3 x 1/157 x 1/100 x 1.85E-03 x 1825 x e -ts.est-12i t:21 a y n is hus.y> )

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Ap = 1.10E-04 Ci

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Dru as = A, x DCF x BR x x/Qus where: Ap = 1.10E-04 (Ci)  ;

DCF = 5.542E+06 (rem /Ci-inhaled) j BR = 3.47E-04 (m*3/sec)  !

x/Que = 1. 08E-03 (sec/m*3) i i

Dru as = 1.10E-04 x 5.542E+06 x 3.47E-04 x 1.08E-03 1 Dru exe = 2. 2 8E- 04 (rem / assembly) (assuming DF= 100]

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Cat. 1 Calc. #

XX-XXX-60RA, Rev. 1  :

i page 14 of 28 1

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Dra ex3 =

2.28E-02 (rem / assembly) [ assuming DF = 1 ) .

For the LPZ, from equation (2) : l 1

I Dra t,: = A, x DCF. x BR x x/Qtp l l

Where: x/Qtp: =

8.35E-05 (sec/m'3) all other parameters are previously defined and calculated l Dra t,2 = 1.10E-04 x 5.542E+06 x 3.47E-04 x 8.35E-05

. Dra t, = 1.77E-05 (rem / assembly) '[ assuming DF=100]

Dru 2,,, = 1.77E-03 (rem / assembly) (assuming DF=1]

l 2) Whole Body Dose from Kr-85 From equation 6, i

A, = Pr x f, x 1/N 3 x Ast xr-es X PL x e' ' *"

_ where: Pr = 1. 7

! .. f, = 0. 3 l l Nx = 157 assemblies i

Asr rr-e s = 4 .102E+ 02 x 2 = 8 2 0. 4 (Ci/Mwt)

P2 , = 1825 (Mwt)

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A = 7.362E-06 (1/hr) l l .- t = 221 (days) I f

i l A, = 1.7 x 0.3 x 1/157 x 8.204E+02 x 1825 x e' .

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Cat. 1 Calc. #

, XX-XXX-60RA, Rev. 1 page 15 of 28 1

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A, - 4 6 7 7 . 4 (Ci)

From equation 4, Dw, ti, = A, x DCF x x/QtAs where: A, = 4 677. 4 i

i DCF = 5.102E-04 (rem-m^ 3 ) / (Ci-sec) l x/Q = 1.08E-03 (sec/m^3) l Dw,ex, = 4677.4 x 5.102E-04 x 1.08E-03 l Dwn cAs = 2 . 5 8 E- 0 3 (rem / assembly) l l

From equation 5 we can determine the LPZ dose. I Dw, t, = A, x DCF x x/Qt, I I

where: x/Qt,2= 8.35E-5 (sec/m^3) all other parameters as defined or calculated above i

Dws trz = 4 677.4 x 5.102E-04 x 8.35E-05 Dw,t,2 = 2.00E-04 (rem / assembly)

3) Skin Dose From KR-85 i

From equation 7, l

Dsr cAs = Dw, ga, x DCFsg a. 3/DCFn_,3 where: Dw, u, = 2 . 5 8 E - 0 3 rem 9

_ m . _._.. _ _._. . - . - _ _ _ ._.

Cat. 1 Calc. #

XX.-XXX-60RA, Rev. 1 page 16 of 28 DCFsr n-ss = 4. 24 6E-02 (rem-m^3 ) / (Ci-sec)

All other parameters as previously '

defined or calculated Dsr us = 2.58E-03 x 4.246E-02/5.102E-04 Dsr us = 2.15E-01 (rem / assembly)

From equation 8, we can determine the Low Population t

Zone beta dose:

l D3 x t, = Dw, t,2 x DCFsr xa-es/DCFa.e3 where: All parameters are as previously defined or calculated.

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D n t, = 2.00E-04 x 4.246E-02/5.102E-04 l 1

D3 g t,2 = 1.66E-02 (rem / assembly) l l

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B) Doses From Cask Drop' j 1) Thyroid Dose From I-129 l From equation (7), we can calculate the thyroid dose i at the Exclusion Area Boundary and Low Population Zone.

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Dn unc... = Dru us x Ra l' where: Dru us = 2 . 2 8 E- 04 (rem / assembly) l

( Ra = 157 (assemblies)

D;n us e.,n = 2.28E-04 x 157 l

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Cat. 1 Calc. #

XX-XXX-60RA, Rev. 1 page 17 of 28 1

Dru gxi c ,, = 3.58E-02 (rem), (assuming DF = 100)

Dru cx3 c ,i = 3.58E+00 (rem) (assuming DF = 1 )

From equation 8, we can calculate the thyroid dose from a. cask drop accident in the spent fuel pool at the Low Population Zone.

Dru tra casa = Dra t, x Ra i

where: Drutrz= 1.77E-05 (rem / assembly) i Rx = 157 assemblies-l Dru t,g c.,a = 1.77E-05 x 157 D;g t,2 c.,g = 2.78E-03 (rem) (assuming DF = 100) i 1

Dru ter e.,a = 2.78E-01 (rem) (assuming DF = 1 )

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2) Whole Body Dose From Kr-85 We can calculate the whole body dose from a cask i i

drop accident in the spent fuel pool using equations 11 and 12. For the Exclusion Area Boundary equation l

10 gives us.

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Dwe tAs c.,x = Dw, gxe x Rx i

where: Dwn ex. = 2.58E-03 (rem / assembly)

Rx = 157 assemblies 4

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Cat. 1 Calc, # -

XX-XXX-60RA, Rev. 1 page 18 of 28 l

1 Dws r.As cask = 4.05E-01 (rem)

For the Low Population Zone, equation 10 gives us: 1 I

bBLFZCask "OWB LP2 x R, where: Dws tre = 2. 00E-04 (rem / assembly)

Rx = 157 (assemblies)

Dwi t, e,,x = 2.00E-04 x 157 1

l Dw, t, e.,a = 3.14E-02 -(rem) 1 l

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XX-XXX-60RA, Rev. 1 page 19 of 28 i

3) Skin Dose From Kr-85 i

l We can calculate.the skin dose for the Exclusion Area Boundary and the Low Population Zone from equations 13 and 14. For the Exclusion Area Boundary, equation 13 is:

Dsr cast ua = Dws casr un x DCFsr Kr-es/DCFx,.,3 where: All parameters are as previously defined and calculated.

D3x .c.,, u3 =

4.05E-01 x 4.246E-02/5.102E-04 Dsrc s= ua = 3.37E+01 rem For the Low Population Zone, equation 14 is:

D3x e.,,  ;,2 = D wn c. ,x ty: x DCF 3r y,.,3/DCFx,.e3 where: all parameters are as previously defined and calculated.

Dsr cask trz = 3.14E-02 x 4. 24 6E-02/5.102E-04 Dsr casa tez = 2. 61 rem 1

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XX-XXX-60RA, Rev. 1 l

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Table 1 Assumptions Assumotion Basis

1) Reactor Power Level = 1825 Mwt Ref. 2, page 3
2) Number of fuel assemblies damaged = 157 Conservatively assumed to be 1 reactor core of 157 assemblies (ref. 3, ^

page 4.1-1) even though 132 assemblies were  :

calculated to be  !

damaged per ref. 1.

3) Fraction of fuel rod activity ref. 4, page 2 which is in the gap:

I-129 = 0.3 Kr-85 = 0.3

4) Specific Activity: ref. 5, library I-129 = 9.26E-04 (Ci/Mwt)

Kr-85 = 4.102E+02 (Ci/Mwt)

[ Note: These values are multiplied by a factor of two in calcu-lation section.)

5) Spent Fuel Pool Decontamination ref. 4, page 2 Factor (DF) For: I-129 = 100

= 1 (conservative)  !

6) Peaking Factor = 1.7 Ref. 10, Table 5.1-1
7) Dose Conversion Factors (DCF): )

I-129m = 5.542E+06 (rem /Ci-inhaled) Ref. 5, Library j Kr-85w, = 5.102E-04 (rem-m'3)/(Ci-sec) Ref. 9, page 21 1 Kr- 8 5, = 4. 24 6E- 02 (rem-m"3 ) / (Ci-sec) Ref. 9, page 21

8) Breathing Rate = 3.47e-04 (m^3/sec) ref.4, page 5
9) x/Q's (sec/m"3 ) : ref. 7, page 13  ;

l EAB = 1.08E-03 LPZ = 8.35E-05 )

[ , 10) Unfiltered Ground Level Release conservative assumption  !

i Cat. 1 Calc. # i XX-XXX-60RA, Rev. 1-page'21 of 28 I

Assumptions (continued)  !

Assumotion Basis l

11) Decay. Constants: Ref. 5, Library  !

Ar-u, = 1.4 0E-15/sec = 5. 04E-12/hr Ag,.n = 2.045E-09/sec = 7.362E-06/hr

12) Reactor Shutdown Date = July 22, 1996 Ref. 6, page 4 i l
13) Maximum Assembly Burnup = 42,955.46 Ref. 6, page C2 '

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14) Number.of Days Fuel is Decayed = 221 See Table 2 of this l Calculation 1

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XX-XXX-60RA, Rev. 1 page 22 of 28 Table 2 Calculation of Fuel Decay Time Month Decay Time (days)

July tu 9 August 31 September 30 October 31 November 30 l December- 31 January (23 31 .

February 28 Sum 221 l

Notes: tu Plant shutdown on July 22, 1996 (ref. 6, page 4).

"' No cask. movement assumed until after. February 28, 1997 assumed.

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Cat. 1 Calc. #

XX-XXX-60RA, Rev. 1 page 23 of 28 Table 3 Reviewers Comments Method of Review: Full Review Performed by Reviewer Reviewer's Comments

1) Page 12 minus sign missing in exponential term.
2) Should we add the Whole Body dose equivalent from the skin dose ?

Precarer's_.Resoonse

1) I agree. Minus sign added to exponential term.
2) No - The NRC does not require beta doses to be calculated and they do not have a beta dose limit for this accident.

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XX-XXX-60RA, Rev. 1 l

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l l VIII) Comments by CY Staff The following comments were received from the CY Staff on Calculation number xx-xxx-60RA, Rev. 1, Radiological Assessment of a Spent Fuel Shipping Cask Drop in the CY Spent Fuel Pool. These comments were transmitted per ref. 13. Following these comments are the i preparer's responses.

1. Q: Why is ICRP 2 methodology used in lieu of ICRP 30 methodology ? )

The calculation should include TEDE and Thyroid CDE calculations i for comparison to EPA PAGs. Please see the attached spread sheets that recalculate doses for the one Assembly fuel damage at the Exclusion Area Boundary. The attached also shows calculations for resulting TEDEs and Thyroid CDEs using guidance provided by EPA 400. 1 A: ICRP 2 dose conversion factors were the original one's accepted  !

by the NRC and the one's used in XX-XXX-60RA, Rev. O. Since the l DCF's from ICRP 2 produced acceptable Exclusion Area Boundary '

Doses it was decided to continue to use them. If the doses were close to the NRC limits for this accident, the DCFs from ICRP 1 30 would have been used. '

Design Basis Accident calculations never have included TEDE I and thyroid CDE calculations for comparison to PAGs. This '

present calculation will be the basis of a future assess-ment which will calculate the TEDE and thyroid CDE in order 1 to provide a comparison to the PAGs.

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2. Q: For the purpose of clarity, showing the resulting doses at different time periods would demonstrate that radioactive decay is not significant for the assumed source term.

A: A note was added to the " Summary of Results" section to show that radioactive decay is not important.

3. Q: The basis for the number of fuel assemblies damaged is not provided. The reference is simply the number of fuel assemblies contained in one reactor core. What is the basis for using one reactor core ? We suggest using the 132 fuel assemblies that is documented in Holtec Report HI-951308, " Analysis of Postulated Cask Drops Over the Connecticut-Yankee Spent Fuel Storage Pool" A: The cover letter (ref. 1) to the Holtec report HI-951308, stated

! "To be conservative, please increase this number- (the 132 l assemblies) by 10 % to 145 assemblies." Since the 145 assemblies t

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l Cat. 1 Calc. # l l XX-XXX-60RA, Rev. 1  !

l l page 25 of 28 is only 12 assemblies short of a full core, I decided to use 157 assemblies in the analysis. In performing these types of calculations over the past twenty years, it is common to have l l

assumptions change. In this case, because the calculated number of 132 assemblies is not cast in stone could change under new assumptions, I thought it would give an added degree of conservatism to use one full core. This may help avoid the l need for future reanalysis.

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4. Q: The average burnup assumption is conservative. Instead of ratio-ing the largest assembly burneo in the spent fuel pool to the J average assembly burnup in the TACT Library, a less conservative approach would be to ratio the averace assembly burnup in the spent fuel pool to the average assembly burnup in the TACT Library.

A: The less conservative approach of ratioing average assembly burnup in the spent fuel pool to average assembly burnup in the TACT Library would be appropriate if we could guarantee that the average burnup of the assemblies damaged by the cask were equal to the average assembly burnup in the spent fuel pool. This of course is not the case. There may be times where a cask could fall in an area of the pool where the average assembly burnup would be less than average assembly burnup in the spent fuel pool ,

and at other times where it would be higher. By using the ratio l of largest assembly burnup to average assembly burnup in the TACT library, we are effectively bounding the activity released.

5. Q: The calculation does not provide the period of time that the 4 scenario would occur over.

A: On page 6 the calculation states under Assumotions: "The release was assumed to be an unfiltered, ground level, instantan.eous release from the fuel building at 221 days after shutdown."

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6. Q: Per CY EPIps, the PARS place the plant in a Site Area Emergency l based on the doses resulting from a cask drop. The calculation '

should be enhanced to include an argument that this scenario can-not escalate to a General Emergency and does not require protective actions. Therefore, bringing the Emergency Plan "to the site boundary" is acceptable.

A: Design Basis Accident Radiological calculations never discuss Emergency Planning PARS. The purpose of Design Basis accidents is to determine if the whole body and thyroid doses meet the requirements of 10CFR100. These discussions will be provided in the justifications for EPlan changes which can reference this calculation.

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XX-XXX-60RA, Rev. 1 page 26 of 28 IX) Comments by Independent Reviewer The following comments were received from an independent reviewer (ref. 12) on Calculation number xx-xxx-60RA, Rev. 1, Radiological Assessment of a Spent Fuel Shipping Cask Drop in the CY Spent Fuel Pool. Following these comments are the preparer's responses.

1. Q: On page 4 of 23, the statement is made that "This analysis assumes one full core. of assemblies are damaged (this is conservative because an analysis shows that a maximum of 132 assemblies could actually be damaged)." "An analysis" should be more specific with reference to a calculation included in the reference section of the calculation.

A: I agree with this comment. I have included a specific reference and refer to it in place of "an analysis".

2. Q: On page 5 of 23 the analysis states that the calculated doses "are under Standard Review Plan,15.7.5 limits of 75 rem thyroid and 6 rem whole body and are therefore acceptable." The dose limits are correct and appropriate; however, I believe that the correct reference would be to SRP 15.7.4 (Reference 3) since SRP 15.7.5 (Reference 4) is applicable to only damaged assemblies contained within the shipping cask and not to fuel assemblies contained within the spent fuel pool that are damaged by the dropped cask. The analysis does conservatively address the release of activity from 157 (one core) assemblies compared to the apparent calculated value of 132 damaged assemblies; however, whether or not this is intended to include assemblies contained within the cask is not specified in the analysis.

A: I agree with the comment on the Standard Review Plan number. l I have corrected the reference to SRP 15.7.4. .

The use of 157 assemblies instead of 132 was meant to give some margin between the calculated value and the value used in the radiological analysis. Based on past experience, the first calculation of some input parameters always seem to be revised upward. The statement in the question regarding if the 157 number included the assemblies in the cask is an excellent example of how the 132 number may increase. The 157 assembly number was therefore meant to give a margin between calculated and that used in the radiological calculation. It does not specify how the number increases, be it by recalculation of the damaged spent fuel pool assemblies, by adding the assemblies in the cask, or a combination of both.

3. Q: On page 6 of 23 the wording for Reference 9 is apparently mis-typed and not understandable.

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XX-XXX-60RA, Rev. 1 page 27 of 28 A: I agree with this comment and have corrected the typo.

4. Q: On page 12 of 23, Equation A is mistyped and should contain (1/DF) instead of (DF).

l A: I agree with the comment and have corrected the typo in the  !

l equation.  !

l t 5. Q: General comment: Addressing the offsite doses from the stand-

! point of only I-129 (thyroid dose contributor) and Kr-85 (whole body and skin) is understandable due to the 179 day decay period. It should be noted, however, that I-131 although l

decayed considerably, is significantly more abundant initially l in the fuel assemblies and is a comparable dose contributor to l

I-129 at the 179 day evaluation period. Although the thyroid dose is somewhat underestimated using I-129 only, the overall results are still considerably below Part 100 limits.

l A: A quick calculation of the I-131 dose contribution shows that

! with 179 days decay, the I-131 dose contribution is about 20% of the total dose. In order to decrease the percentage contribution from I-131, I have extended the decay time to 221 days. Using l 221 days decay adds about 5 half-lives to the decay of I-131 and l results in an I-131 dose contribution of only .65% of the total l

thyroid dose. The calculation showing this is included as Attachment A. Since by the time we actually move fuel or casks the I-131 should have decayed completely, the I-131 is not included in the main calculation.

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XX-XXX-60RA, Rev. 1 page 28 of 28 I

O X) Attachments 1

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XX-XXX-60RA, Rev. 1 pace'Al of A2 Attachment A 1

I-131 Dose Contribution l 179 Devs Decay l

Ap = P x f, x 1/N, x 1/Dr x Asp I-131 X PL X e' **

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l where: Pf = 1.7 f, = 0 .1 i

N3 = 157 assemblies Df =1 A ,, 2 131 = 2. 5 08E+ 04 (Ci/Mwt)

P,= 1825 (Mwt) 3 A = 3.586E-03 -(1/Hr) t = 179 Days Ap = 1. 7 x 0.1 x .1/157 x 1 x 2. 508E+04 x 1825 x e"8 ""-033 8DH i20 Ap = 0.0101 (Ci)

Drx u, = Ap x DCF x BR x x/Que Where: Ap = 0.0101 (Ci)

DCF = 1.4 9E+06 - (Rem /Ci-Inhaled)

(All other parameters as defined previously)

Dru u, = 0 . 0101 x 1. 4 9E+ 0 6 x 3 . 4 7E- 04 x 1. 0 8 E- 03 i

Dra u, = 5. 6E (Rem) l 221 Days Decay l

A, = P, x f, x 1/N3 x 1/Dr x A ,,3 232 x Pr, x e*At I

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Cat. 1 Calc. #

. XX-XXX-60RA, Rev. 1

  • page A2 of A2 Where: t = 221 (days) {

(all other parameters same as above) ,

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A, = 1. 7 x 0 .1 x 1/157 x 1 x 2 '. 5 0 8 E+ 04 x 182 5 x e ns,ssst.03)(22n tro A, = 2 . 7 2 E - 0 4 (Ci) j Drn zu = Ap x DCF x BR x x/Que

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where: A, = 2. 72E-04 (Ci)

(all other parameters are as defined previously) 1 l

l Dru as = 2.72E-04 x 1.4 9E+06 x 3.47E-04 x 1. 08E l Drucas = 1.5E-04 (rem)  !

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A7,techment B page B1 of 1 ,

CTP DAT1 BASE INPUTS l Calculation Number:

Date 3 /of/ 1# 1 (prefix) (sequence (suffix) Revision t number)

Vendor Calculation Number /Other:

O~N' N QA O Yes O No Supersedes

'CCN#

Cale: WN ~ j lWS i Superseded By: l li EWR Number Component Id Computer Code Rev.

Unit

  1. / Level CT NA NA nono )

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PMMS CODES

  • l .

Structure System Component '

i XX XXX - XXX l V6 F MV - DX ,

l-. i I

'The cedes required must be the alpha codes designated for structure, system and component l

.l1 Comments:  !

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i Reference Drawings Sheet i Reference C=lentation 1

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1 NUC DCM FORM 5-1B Rev.04 l 1

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l Attachmant C page C1 of C1 Eems nEv. 54 , TELEPHONE MEMORANDUM r,.OJECT

' ROUTING CY Cask Drop Accident NAME hNmALS CALL oATE TIME X AM 01/21/97 09:30 pu iueouiNo C ouTooivo l TELEPHONE CONVEREATION PARTICIPANTS j l

NAME COMPANY NAME l l )

Donald Miller N11SCo Jay Lee NRC l

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l CALL REPORTER's NAME DATE Donald Miller 01/21/97 i cc: Chronological File

SUMMARY

I i I l I initiated the call to ask Jay the following question: Upon what burnup are

! i l the specific activities in the TACT Library based ? l 1

Jay said that people at the NRC were asking him the same cuestion. He said

, that the highest value was 40,000 MWD /MT and the average was 23,000 MWD /Mt.

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