B14667, Forwards Addl Info Re 931027 Amend Request Re Suppl Leak Collection & Release Sys.Pages from Calculation 88-019-96RA Also Encl

Forwards Addl Info Re 931027 Amend Request Re Suppl Leak Collection & Release Sys.Pages from Calculation 88-019-96RA Also Encl

ML20059F850
Person / Time
Site:	Millstone
Issue date:	10/29/1993
From:	Opeka J NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
B14667, NUDOCS 9311050117
Download: ML20059F850 (17)

Text

I

., - i NORTHEAST UTILIT9ES ceners Omca . semen street, sernn. Connecucui i l umm. mesnu xw '*" P.O. BOX 270

.r.u w n w,a w w.

HARTFORD. CONNECTICUT 06141-0270  :

k L J CZZZ (203) 665-5000 l

October 29, 1993 l Docket No. 50-423 >

B14667 i Re: 10CFR50.90  !

U.S. Nuclear Regulatnry Comission Attention: Document Control Desk -I Washington, DC 20555 i Gentlemen:

Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specification Supplementary Leak Collection and Release System Reauest for Additional Information In a letter dated October 27, 1993,"' Northeast Nuclear Energy Company ,

(NNECO) proposed a license amendment to its operating license, NPF-49, by  :

incorporating the- changes to the Millstone Unit No. 3 Technical l Specifications. The purpose of the submittal was to consolidate the proposed i license amendments dated July 29, 1993,t2' and October 22, 1993, into a l single proposed license amendment to ensure continuity. The submittal also incorporated additional information requested by the NRC Staff during a  :

conference call conducted on October 19, 1993, and during a meeting conducted on October 25, 1993. The proposed changes principally concern the Millstone  :

Unit No. 3 Technical Specifications related to the supplementary . leak i collection and release system and auxiliary building filter system. In -

addition, in the October 27, 1993, submittal, NNECO requested that the NRC ,

Staff process the license amendment request on an emergency basis pursuant to i 10CFR50.91(a)(5).

In a telephone conversation on October 29, 1993, the NRC Staff requested j additional information concerning the assumptions used in the post loss-of-- l t

(1) J. F. Opeka letter to the U.S. Nuclear Regulatory Comission, " Proposed Revision to Technical Specifications, Supplementary Leak Collection and  ;

Release System," dated October 27, 1993. l (2) J. F. Opeka letter to the U.S. Nuclear Regulatory Comission, " Proposed  !

Revision to Technical Specifications, Supplementary Leak Collection and j Release System," dated July 29, 1993.  !

(3) J. F. Opeka letter to the U.S. Nuclear Regulatory Comission, " Proposed i Revision to Technical Specifications, Supplementary Leak Collection and i Release System," dated October 22, 1993. .

. 9311050117 931029 C [

os3c2 nty oe Qlj PDR ADOCK 05000423 h t

• k P PDR I

- \

i U.S." Nuclear Regulatory Commission B14667/Page 2 [

October 29, 1993 t l

I coolant accident dose assessment presented in our October 27, 1993, submittal.  ;

The requested information is provided in Attachments 1 and 2.  ;

We apologize for the confusion introduced by the incomplete updating of the f

, Millstone Unit No. 3 FSAR following issuance of Amendment No. 59 associated i i with the change of containment pressure. Correction of this situation will be  !

undertaken promptly. A reconciliation is provided in the Attachments.  !

1 If you have any questions regarding this submittal, please contact our  !

licensing representative directly. l Very truly yours, j NORTHEAST NUCLEAR ENERGY COMPANY  !

I FOR: J. F. Opeka l

Executive Vice President  !

i BY: A M u[+ l E. A. DeBarba *

Vice President  ;

cc: T. T. Martin, Region I Administrator 4 V. L. Rooney, NRC Project Manager, Millstone Unit No. 3 l

'. P. D. Swetland, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 .

1' I Mr. Kevin T.A. McCarthy, Director Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street  ;

P.O. Box 5066 j Hartford, CT 06102-5066 ,

Subscribed and sworn to before me l this o & day of v/ b /e M , 1993 4

^

T %_.[ v 2

Date Commission Expires: # / 6 f

Katnton T. Garc i 9

~

tbtnrv PubCc lh C&~s.cn Eerh & ,

t t

I

f Docket No. 50-423  :

B14667 i I

.i i

Attachments 1 and 2 '

Millstone Nuclear Power Station, Unit No. 3 f Additional Information Related to the  !

Amendment Request dated October 27, 1993 l, t

i r

L l

i l

f Sctober 1993 i

. U.S. Nuclear Regulatory Commission B14667/ Attachment 1/Page 1 October 29, 1993 Attachment 1 Resoonses to Reouest for Additional Information

1. NNECO has confirmed that our calculation starts with 50% of the core '

inventory of iodine in the containment air, and instead of assuming an instantaneous plate-out of 50% of that iodine, we assumed zero plate-out at t=0 and assumed a plate-out removal factor of A = 0.176/hr for elemental iodine (see pages 13,14,14a of Enclosure 1 to this letter). .

2. The manner in which we took credit for the spray is shown in the attached pages from calculation #88-019-96RA (see pages 6, 7, 8, 9, 12, and 21 of Enclosure 1 to this letter).

3. The following paragraph is quoted from Attachment I to the NNECO letter dated December 6, 1990.  ;

Pace 5 of Attachment 1. item 4 "4. ComDutation of Fission Product Removal Rate SRP 6.5.2 states in part that acceptable methods for computing fission product removal rates by the spray system are given in Subsection 111.4.C of SRP 6.5.2, ' Fission Product Cleanup Models.'

For Hillstone Unit No. 3, calculation of the elemental iodine removal coefficient is based on the model presented in the ANSI /ANS  !

56.5-1979, 'PWR and BWR Containment Spray System Design Criteria,'

rather than that presented in Subsection III.4.C of SRP 6.5.2. The ,

model specified in the ANS standard considers the effect of pH on l the iodine removal coefficient. The model specified in the SRP does '

not consider this effect with the exception that an upper limit is placed on the value for boric acid spray. For the spray droplet diameters and pH values applicable to Hillstone Unit No. 3, the ANS standard is more conservative than the SRP."

i l

i

[

U.S. Nuclear Regulatory Commission B14667/ Attachment 2/Page 1 October 29, 1993 i

Attachment 2 Comparison between the Millstone Unit No. 3 FSAR 7 Table 15.6-9 and NNEC0's Submittal dated February 26. 1990 '

> TABLE 15.6-9 ASSUMPTIONS USED FOR THE LOCA February 26, 1990 '

ANALYSIS Submittal i

Expected Design Power level (MWt) 3,636(1) 3,636 Not specifically included. t Operating time (days) 650 650 Not specifically included.

Fraction of fuel defects 0.005 0.01 Not specifically included.

Core inventory Table 15.0-7 Table 15.0-7 Not specifically included. ;

Iodine composition Elemental (%) 95.5 95.5 95.5 (Page 14)

, Particulate (%) 4.0 4.0 2.5 (Page 14)

Organic (%) 0.5 0.5 2.0 (Page 14)

Fraction of core inventory released into reactor coolant Iodine 0.02 0.5 Not specifically included.

• Noble gas 0.02 1.0 Not specifically included.

Fraction of reactor Not specifically included.

l coolant inventory However, for iodine we ,

used 1.0 in our d

available for release from containment calculations. Therefore, Iodine 0.5 0.5 FSAR Table does not 4

Noble gas 1.0 1.0 represent our assumptions correctly.

Core inventory, available Not specifically included.

for release from However, for iodine we containment used 50 % in our Iodine (%) 1.0 25 calculations. Therefore, Noble gas (%) 2.0 100 FSAR Table 15.6-9 does not represent our assumptions ,

correctly. ,

Containment free volume 2.3 x 10' 2.3 x 10' Not specifically included. l (ft') l 4

i

U.S. Nuclear Regulatory Commission '

B14667/ Attachment 2/Page 2 October 29, 1993 TABLE 15.6-9

, ASSUMPTIONS USED FOR THE LOCA February 26, 1990 ANALYSIS Submittal Containment leak rate (percent per day) i 0-24 hr 0.65 0.65 Page 12 of the submittal.24-720 hr 0.325 0.325 Page 12 of the submittal.

Bypass leakage (fraction 0.04277 0.04277 Page 15 of the submittal.

of containment leakage) ,,

Elemental iodine plate-out Page 16 of the submittcl rate 0.176/hr 0.176/hr says it is calculated in -

the model. t l

Containment spray assumptions:

<

• Volume of sprayed region - 1.20' x 10 8 ft 8

• Volume of unsprayed region - 1.114 x 10 ft Iodine decontamination factor (DF) assumed Not included in our ,

during quench spray operation - 200 submittal. However, our  !

calculations allowed --  !

200, actual DF based on  !

TACT 3 calculations using t As.

• Iodine DF during recirculation spray Not included in our l operation - 12 submittal. However, our i calculations assume as follows: Maximum allowed-- '

12 reached in 1.5 hrs using As.  ;

• Quench spray operation initiation time - 0

• Recirculation spray initiation time - 750 sec.

• Mixing rate between sprayed and unsprayed region - 2 turnovers /hr

• Iodine removal rates in spray region:

, A elm - 28.1/hr A part - 2.16/hr +

Duration of release from 720 720 Page 13 of the submittal.

containment (hr)

'f a

.  ?

l U.S. Nuclear Regulatory Commission i B14667/ Attachment 2/Page 3 i October 29, 1993 TABLE 15.6-9  !

ASSUMPTIONS USED FOR THE LOCA February 26, 1990  !

ANALYSIS Submittal  !

Post-LOCA Equipment Leakage  !

4 Leakage initiation and (2) 220 sec to Not specifically included cessation times 720 hr in the submittal.  !

Maximum operational leak (2) 5,000(4) Not specifically included  !

rate (cc/hr) in the submittal. j Fraction of core iodine (2) 0.50 Not specifically included ,

inventory in sump water in the submittal. i Sump water temperature (2) 256-<212 Not specifically included

(*F) in the submittal.

Iodine released to (2) 10(3) Not specifically included  !

building atmosphere from in the submittal. l l recirculation leakage (%) j Figure 4 of the submittal.

Filter efficiency  ;

Elemental iodine (%) 95 95 Methyl iodine (%) 95 95  !

HEPA (%) 95 95 i I

NOTES:

1. Includes instrument error of 2 percent '

2. Not applicable. ,

3. Despite temperature variation, at no time is there a Not applicable, j greater than 10 percent of the water in the sump '

flashing steam.

4. To be conservative, the calculation assumed the  ;

maximum post-LOCA equipment leakage was a factor of two times the max operational leakage to give a total ,

leakage of 10,000 cc/hr. ,

1 i

i e

ENCLOSURE 1 EAB AND LPZ DOSES FROM A UNIT 3 LOCA  !

QA CATEGORY 1 88-019-96RA l 21 Pages l i

. , l i

l  !

Method of Review:

Full Review Performed by Reviewer Preparer: IIl4!E7 gM Wheeler  :

Reviewer: L . I al 73 D. W. Miller

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Approved by: O. W ,

f-ao sf p R. A.Crandall

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TACT III FEB. 83 VEpsIO4 = ACCIDENT AHtLYSIS BEAUCH CAPD IttPUT CASE IIO. 1 ,

TIME It0EFENCEtiT It1PUT .

00000100 MP-3 LDC A ANALYSIS ( 0.027BX/D AY UHFILTERED) 50% REDUCTIC14 AT 24 HOURS '

2 8 2 21 1 16 89 3.636E+03 0.0 5.000E-01 1.000E+00 1.000E*C0 9.550E-01 2.000E-02 2.500E-02 >

TIME DEFENDENT It4PUT ,

VALUES RE AD IN FOR X App AY  !

ISET IDATA IHUN l

1 2 0 0.0 2.083E-01  !

2 2 0 5.200E-01 4.800E-01 1.114E+06 [

3 2 0 1.20 tee 06 5 2 0 2.8 8E*01 0.0 7 2 0 2.160E*00 0.0 3.712E*04 10 3 1 0.0 0.0 10 3 2 0.0 3.71EE*04 0.0 11 3 1 2.780E-02 0.0 0.0 j

11 3 2 2.780E-02 0.0 0.0 2.910E-05 3.470E-04 0.0 0.0 2.780E-02  ;

17 6 0 5.420E-04 i

4 1 2 0 2.083E-01 1.000E+00 1 2 0 1 000E*00 1.533E+00 1

1 2 0 1.533E+00 2.000E*00 .

5 2 0 0.0 0.0 I 2 0 2.000E*00 8.000E*00 2.910E-05 3.470E-04 0.0 0.0 2.780E-02 t 17 6 0 0.0 2.400E*01 l 1 2 0 8.000E+00 2.780E-02 1.990E-05 1.75CE-04 0.0 0.0 l 17 6 0 0.0

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I 1 2 0 2.400E*01 9.600E*01 t 3 1.390E-02 0.0 0.0 '

11 1 3 2 1.390E-02 0.0 0.0 11 0.0 1.390E-02 17 6 0 0.0 8.660E-06 2.320E-04 0.0

[

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2.630E-06 2.320E-04 0.0 0.0 1.390E-0*

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• ACCIDENT AMALYSIS BRANCH '

CARD IMPUT

} CASE NO. 1 TIME INDEPENDENT INPUT

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88000*410 MP-5 LDCA ANALYSIS TOTAL BYPASS Tc 2.5 MIN 1587. I, 200'./ NG) '

2 8 2 21 le 14 95 5.000E-81 1.000E+00 1.000E+ce 9.55eE-01 2.000E-02 2.500E-82 [

, 5.656E + 05 0.9

! TIME DEPENDENT INPUT ,

j ISFT 1 DATA INUM VALUES READ IN FOR X ARRAY f r

l 1 2 8 0.0 4.167E-02 2 2 9 5.?00E-01 4.800E-01 5 2 e i.20dE+0e 1.114E*06 l 5 2 0 2.826F

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5 2 8.0 5. 712E + D4 0.9  ;

le 0.0 11 3 1 3.000E-01 0.8 i 2 3.00sE-01 0.0 0.8 11 5 2.910E-85 3.476E-64 0.0 0.0 2.780E-82 17 6 0 5.420E-D4 f

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