ML20115A873

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Rev 2 to Shutdown Cooling Sys Vol & Refueling B Concentration Calculation
ML20115A873
Person / Time
Site: Millstone Dominion icon.png
Issue date: 06/28/1996
From: Borchert R, Honan T
NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20115A866 List:
References
96-ENG-1392-M2, 96-ENG-1392-M2-R02, 96-ENG-1392-M2-R2, NUDOCS 9607090060
Download: ML20115A873 (31)
v • d • e


Text

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%ENG-4392-m2:

2. Y.

CALCULATION #

REV. # -

System _ 9pc./RCL

_ Structure 4/4 Component N/#

l Executive Summary

(2)

Estunate the volume of boric acid solution required to increase the SDC/RCS boron concentration from 1450 ppm to 2100 ppm.

(3)

Determine the minimum SDC boron concentration required to ensure acceptable results in the event of inadvertant mixing with water which has a lower boroa concentration from the stagnant portions of the RCS.

  • This
  1. calculation 2-5-96. is being performed to support a change to Proposed Technical Specification Change Req The results of this calculation are:

(1) The total estimated SDC water volume is 34,623 gallons.

(2) 6,619 gallons of boric acid at 5500 ppm are required to increase the SDC boron concentration from 1450 ppm to 2100 ppm.

(3)

Approximately 17,731 gallons of" stagnant" water will remain in the RCS following draindown of the RCS to 3 feet below the reactor vessel flange.

(4)

A boron concentration value of 1950 ppm is sufTicient to satisfy Technical Specification LCO 3.9.1 (assuming that all " stagnant" RCS water mixes with the SDC water).

Does this calculation:

1.

Support a DCR, MMOD, an independent review method for a DCR ,or confirm ntest A Yes ONo results for an installed DCR7 Ifyes, indicate the DCR, MMOD number and/or Test Procedure number.

2.

_3 itSupport supports. independent analysis? Ifyes, indicate the procedure or work control reference Yes ONo @

/

3.

Revise or supersedes, or voids existing calculations? Ifyes, indicate the calculation cumber and revisions.

4. W GJr -1T12-M2 . QeJ. t Yesho O Involve QA-related systems, components or structtires (ADVS, Fire Protection, Radwaste. FM R/hm n)? -

Yes oO Approvals Preparer /Date:

_ . l A.be 4 4. @cA4 /MM g/M i

Independent Reviewer /Date: 4,/27/96

  • Design Engineering Supervisor /Date: %A J.Tod / WMm _

y '

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5. (orm.de /M. ddd

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b/ev/e/4

/

9607090060 960703 NUC DCM FORM 5-1A PDR A00CK 05000336 Rev.02 P PDR

i i

Northeast .

Wlities Systern  ;

i CTP DATA BASE INPUTS .

\

' NUSCO Calculation Number:

' 96-6 4 1392. M7-. Revision 2..

' (prefix) (wussce *'

b /27/96 (suffix) Date l

number)  !

l Vendor CalE1= don NumNr/Other:

W Coobg S b y

. i

Title:

%Le a (2e fan, CCN# ____ SupersededBy:

QA(y/n)

Go% C.,c Arab,- Jb j Unit Pa Number ComponentId Computer Code Rev.

  1. /Ievel mP HlA SM-/VCS N)A NIQ PMMS CODES Stmeture ' System Component Reference Calculation ' Reference Drawings ' Sheet a[A 23cn/23o app 96-T52-I%L -mz 252o1-2139 19 96-F4-/792-m2.

Comments:

NUC DCM FORM 5-1B Rev 00 ,

9

. .. . . .- ~

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+

a Calculation # 96-ENG-1392-M2, Rev. 2 Table of Contents i

1. Purpose 1
2. Summary ofResults ,

1

3. References / Design Inputs '

1

4. Assumptions 2 *
5. Method ofCalculation 3
6. Body ofCalculation 4 '

6.1 Estimate of water Volume in Shutdown Cooling System i

4 6.2 Estimate of SDC Water Volume in Reactor Coolant System 5 6.3 Total Estimated Shutdown Cooling System Water Volume 7 6.4 Estimate of" Stagnant" Water Volume in RCS Following Draindown 7 i 6.5 Volume ofBoric Acid Required to Increase SDC to 2100 ppm 10 6.6 Evaluation ofRequired SDC Boron Concentration 10 i i

7. Reviewer's Comments and Resolution 12
8. Attachments 13 E

6/27/96

Calculation # 96 ENG-1392-M2, Rev. 2 Prepared By: g6[] 6/27/%

Reviewed By:f78 gw/((,

1. PURPOSE:

The purpose of this calculation is to

, (1) Estimate the water volumes in the Millstone Unit 2 Shutdown Cooling -

(SDC) system and the Reactor Coolant System (RCS) when the RCS is drained to 3 feet below the reactor vessel flange.

l  ;

(2) Estimate the volume ofboric acid solution required to increase the SDC/RCS boron concentration from 1450 ppm to 2100 ppm.

(3) Determine the minimum SDC boron concentration required to ensure l acceptable results in the event ofinadvertant mixing with water which has a lower boron concentration from the stagnant portions of the RCS.

This calculation is being performed to support a change to Proposed Technical l

Specification Change Request # 2-5-96.

2.

SUMMARY

OF RESULTS:

/ 2.1 The estimated water volume in the SDC system heat exchangers and piping is 6,925 gallons.

/

2.2 The estimated water volume in the RCS which is actually mixed by the SDC system is 27,698 gallons.

l

/ 2.3 The total estimated SDC water volume is 34,623 gallons. l

,/ 2.4 6,619 gallons of boric acid at 5500 ppm are required to increase the SDC boron concentration from 1450 ppm to 2100 ppm.

[ 2.5 Approximately 17,731 gallons of" stagnant" water will remain in the RCS following draindown of the RCS to 3 feet below the reactor vessel flange.

/ 2.6 A boron concentration value of 1950 ppm is sufficient to satisfy j l

Technical Specification LCO 3.9.1 (assuming that all " stagnant" RCS >

water mixes with the SDC water).

3. REFERENCES / DESIGN INPUTS:

3.1 Calculation number 96-TS2-1356-M2, Rev. O, " Determination of Shutdown Cooling System Piping Linear Footage and Volumes" 4

Page 1 of13

, 6/27/96

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

,  ;. i i

, Calculation # 96-ENG 1392-M2, Rev. 2 -

Prepared By: 7 (-

ReviewedBy:% 6 {

T3W ff-  ;

3.2 Combustion Engineering memo F-MD-9dN-MD-146, dated June 5,1972, ,

" Water Volumes and Flow Areas for Hutchinson Island and Millstone  ;

. , Unit 2 (02-32-00)"(CDCC # 43173) 3.3 Millstone Unit 2 Reactor Coolant Piping Instruction Manual, Table 1.1 and Drawing E-234-000, Rev. 4, " General Arrangement ofPrimary Piping" l

4 .

3.4 NUSCO Drawing No. 2520' 3-29139, Sheet 19, Rev. 5, " General l

Arrangement - Elevation,172" I.D. PWR" '

3.5 Babcock and Wilcox Performance Data Report, BWC-222-7612-PR-1, .  ;

Figure 3.6 3.6 Millstone Unit 2 Final Safety Analysis Report,. Table 4.3-4 i 3.7 Siemens Power Corporation letter TMH:96:075, dated April 2,1996, T.M. Howe to C.H. Wu, " Refueling Boron Concentration for Millstone Unit 2 Cycle 13 Mid-Cycle Shutdown"

4. ASSUMPTIONS:

/ 4.1 The water volume in the reactor vessel .oove the top of the RCS hot leg elevation is not mixed by SDC flow.

4.2 No mixing occurs between SDC and the water volume in the RCS Loop I hot leg (including steam generator and pressurizer).

l

/ 4.3 No mixing occurs between SDC and the water volume in the RCS Loop 2 hot leg downstream of the SDC nozzle (including steam generator).

l

[ 4.4 No mixing occurs between SDC and the water volumes in the RCS cold leg Loops 1 A, IB,2A and 2B upstream of the Safety Injection nozzles (including steam generators and reactor coolant pumps).

/ 4.5 The boron concentration in the stagnant portions of the RCS is at 1300 ppm. This is a conservative value and was approximately the RCS boron concentration when the Reactor Coolant Pumps were stopped in Febmary 1996.

/- 4.6 The boron concentration in the boric acid storage tank is at 5500 ppm. ,

This was the concentration of the "A" BAST on April 27,1996, and can be  ;

considered to be a representative value.

Page 2 of 13 6/27/96

. l 2

Calculation # 96-ENG-1392-M2 Prepared By:pt Rey,6 6)27/% l Reviewed By: Tyd 6/es/f4

[4.7 The water level in the reactor vessel is at 3 feet below the reactor vessel head flange. This is where the water will be drained to perform the SDC i

. , boration from 1450 ppm to 2100 ppm.

/ 4.8 Estimated portion ofLoop 2 hot leg that experiences SDC flow is 70% )

(0.7). Likewise, the estimated portion ofLoop 2 hot leg that does not experience SDC Bowis 30% (0.3). )

/ 4.9 Estimated portion ofLoops I A, IB, 2A and 2B cold legs that experience SDC flow is 70% (0.7). Likewise, the estimated portion of this piping that does not experience SDC flow is 30% (0.3). This only applies to the piping between the reactor coolant pump discharge to the reactor vessel  ;

inlet nozzle.

! .10 4 The water volume in the pressurizer surge line is assumed to be the entire volume of piping assembly No. 505-12. This is a conservative value as this  !

pipe will be partially drained when water level is at 3 feet below the reactor i vessel flange. The horizontal section of the surge line piping is at elevation 11 feet 2% inches (or about 15 inches below the reactor vessel flange).

4.11 The steam generator primary heads are assumed to remain completely filled (due to the complicated calculations involved). This is a conservative assumption, since they will'actually be partially drained when the water level is at 3 feet below the reactor vessel flange.  ;

/ 4.12 The steam generator tubes are assumed to completely drain when the water l level is at 3 feet below the reactor vessel flange. This is a reasonable j assumption, since the bottom of the tubesheet is at elevation 11 feet 8%

inches (or about I foot below the reactor vessel flange), and the tubes had been previously drained in March 1996.

5. METHOD OF CALCULATION:

/ 5.1 The various volumes are obtained from the appropriate reference documents and are summed in order to provide the estimated water volumes.

l

-/' 5.2 The required SDC boron concentration is calculated by subtracting the product the volume and boron concentration of the " stagnant" loops from the product of the sum of the " stagnant" and SDC volumes and the required refueling boron concentration. The resultng value is then divided by the SDC volume.

Page 3 of 13 6/27/96

Calculation # 96-ENG-1392-M2. Rev. 2 Prepared By: % 4 ~2'T %

Reviewed By: 77N 4 -

6. BODY OF CALCULATION:

i

, , , 6.1 Estimate ofWater Volume in Shutdown Cooling System-i

- NOTE: The volumes used in this portion of the calculation were obtained from Reference 3.1.

a. SDC Iniection Header Pioina Volumes Loop 1 A piping volume = 62.52 ft' Loop 1B piping volume = 58.22 ft' /

Loop 2A piping volume is calculated using the sum of several portions from the Reference 3.1 calculation:

1.34 ft' + 2.35 ff +30.9 ff + 9.06 ft' + 5.38 ft' = 49.03 ft' Loop 2B piping volume = 56.23 ft' /

b. SDC Heat Exchanger and Associated Pioine Volumes Water volume of 2 SDC heat exchangers = 176.46 ft' /

Piping volume to and from SDC heat exchangers = 129.23 ft' /

c. SDC Suction Pining Volume SDC Suction Header Volume 28.59 ft' + 29.28 ft' + 12.41 ff + 1.37 ff + 68.89 ft' = 140.54 ft'  !

"A" LPSI Pump Suction Piping Volume 50.07 ff + 19.76 ft' = 69.83 ft' "B" LPSI Pump Suction Piping Volume 40.95 ft' + 17.59 ff = 58.54 ft' Page 4 of 13 6/27/96

Calculation # 96-ENG-1392 M2, Rev. 2 Prepared By: @*6 Ef2 L Reviewed By: TJ74 4/2

d. LPSI Pumo Discharee Pioine Volume

.. "A" LPSI Pump Discharge Piping Volume = 21.56 ft' /

"B" LPSI Pump Discharge Piping Volume = 24.33 ft' /

Discharge to SDC Injection Header Piping Volume 28.75 ft's 50.51 ft' = 79.26 ft'

e. Estimated Water Volume in SDC System Sum of SDC piping volumes = 925.75 ft' /

Conversion factor is 7.48 gallons /ft' Estimated gallons of water in SDC system = 6,925 gallons /

6.2 Estimate of SDC Water Volume in Reactor Coolant System NOTE: This is an estimate of the RCS water volume that is aciually circulated by the SDC system.

a. Reactor Vessel Water Volume to Too of RCS Hot Legs NOTE: The reactor vessel volumes used in this calculation were obtained from Reference 3.2.

Elevation 1 (Lower RV head to core support plate) = 1114 ft' Elevation 2 (Active core region) = 1140 ft' Elevation 3 (Top of core to bottom of hot leg) = 449 ft' ,

I Elevation 4 fRCS Hot Leg region) = 671 ft' /  !

Reactor Vessel Volume = 3374 ft' /

Page 5 of 13 l 6/27/96 j l

Calculation # 96-ENG-1392-M2, Rev. 2 Prepared By: h ReviewedTJN By:&% 6/'

NOTE- The RCS piping volumes were obtained from Refernnce 3.3.

, .. b. RCS Looo 2 Hot Lee to SDC Suction Pioing Volume Volume of Assembly No. 503-02 = 140.6 n' O.7 x 140.6 R' = 98.4 ft' /

c. RCS Looo 1 A Cold Lee to Reactor Vessel Pioine Volume Volume of Assembly No. 504-04 = 82.3 R' !

3 0.7 x 82.3 R = ,

57.6 ft'

d. RCS Looo IB Cold Lee to Reactor Vessel Piping Volume Volume of Assembly No. 504-05 = 82.1 R' /

3 0.7 x 82.1 R = 57.5 ft'

e. RCS Looo 2A Cold Lee to Reactor Vessel Pioine Volume Volume of Assembly No. 504-01 = 82.3 R' /

0.7 x 82.3 R =

57.6 ft' '

f. RCS Looo 2B Cold Lee to Reactor Vessel Pioing Volume Volume of Assembly No. 504-03 = 82.1 R' / l

\

0.7 x 82.1 R =3 57.5 ft' l

g. Estimate of SDC Water Volume in RCS ,

Sum of reactor vessel and RCS piping water volume = 3703 ft' /

i Conversion factor is 7.48 gallons /R' Estimated gallons of SDC water in RCS = 27,698 gallons /

1 i

Page 6 of 13 l 6/27/96

]

Calculation # 96-ENG-1392 M2, Rev. 2 Prepared By: f4d 6  %

Reviewed By: 2TH 6.3 Total Estimated Shutdown Cooling System Water Vdume Estimated gallons of water in SDC system =  ; ,6,925 gallons Estimated nallons of SDC water in RCS = 27.698 nallons [

34,623 gallons / j Total estimated gallons of SDC water =

6.4 Esimate of"Stannant" Water Volume in RCS Following Draindown NOTE- This is an estimate of the water volume in the RCS that is up_t  :

I mixed by SDC flow after the RCS is drained to 3 feet below the '

reactor vessel flange.

. l The reactor vessel water volume is obtained from Reference 3.2, the RCS piping volumes are obtained from Reference 3.3, and the RCP volume is obtained from Reference 3.6.

a. RV Uooer Plenum Region (Too of Hot Lee to RV Flance) i Elevation 5 (Top of Hot Leg to top of RV) = 1296ft' '

- Thermal Region 6 (RV Head) = -673 ft'

- Thermal Region 7 (CEDM Nozzles) = -26f1 3 Total RV Upper Plenum Region Volume = 597 ft' '

From Reference 3.2, the top of Elevation Region 4 is the top of the Hot Leg Nozzle at the ID of the reactor vessel.

From Reference 3.4, the I.D. of the Hot Leg Nozzle at the ID of the reactor vessel is 48.125 inches. Thus, the distance from the centerline of the nozzle to the top of the hot leg is 24.0625 inches.

From Reference 3.4, the distance from the cpterline of the nozzle to the reactor vessel flange is 78 inches Thus, the height of the upper reactor vessel plenum region is calculated as:

78 inches - 24.0625 inches = 53.9375 inches /

Change in RV Upper Plenum Region volume per 1 inch change in RV water levelis:

597 ft' + 53.9375 inches = 11.1 ft'/ inch /

Page 7 of 13 6/27/96

. l e

Calculation # 96-ENG-1392-M2, Rev. 2 I

Prepared By: fC66 94 Reviewed By: 77#

Smce the RV water level has been decreased to 36 inches below the RV flange, the amount of water remaining in the RV upper plenum region is: ,

. 597 ft' -(36 inches x 11.1 ft'/ inch) = 197.4 ft' I

b. RCS Loon 2 Hot Lee (SDC Nozzle to SG 2) Pioine Volume Volume of Assembly No. 503-02 = 140.6 ft' /

0.3 x 140.6 ft' = 42.2 ft' *'

c. RCS Looo 1 Hot Leg Pioina Volume Volume of Assembly No. 503-01 = 143 ft' / l
d. RCS Loco 1 A Cold Lee to Reactor Vessel Pioine Volume NOTE: This includes the piping from steam generator 1 to reactor coolant pump to safety injection nozzle.

Volume of Assembly No. 503-03 = 28 ft' Volume of Assembly No. 505-10 =

53.5 ft'

30.5 ff Volume of Assembly No. 503-05 = i Volume of Assembly No. 504-04 = 82.3 ft' x 0.3 = 24.7 ff '

Volume of reactor coolant numo = 112 ff '  ;

RCS Loop 1 A Cold Leg Piping Volume = 248.7 ft'

e. RCS Loon IB Cold Leg to Reactor Vessel Piping Volume NOTE: This includes the piping from steam generator 1 to reactor coolant pump to safety injection nozzle.

28 ft' Volume of Assembly No. 503-03 =

Volume of Assembly No. 503-07 = 53.5 ff /

Volume of Assembly No. 503-05 = 30.5 ft' Volume of Assembly No. 504-05 = 82.1 ft' x 0.3 = 24.6 ff '

Volume of reactor coolant ounio = 112 ff /

RCS Loop 1B Cold Leg Piping Volume = 248.6 ft' #

Page 8 of 13 6/27/96

Calculation # 96-ENG-1392-M2, Rev. 2 Prepared By: $43b/2 i6 Reviewed By: 77H V2Vh

f. RCS Looo 2A Cold Lee to Reactor Vessel Pining VolurDD NOTE: This includes the piping from steam generator 2 to reactor

' coolant' pump to safety injection nozzle.

~

Volume of Assembly No. 503-03 = 28 ft' / i Volume of Assembly No. 505-10 = 53.5 ft' / '

Volume of Assembly No. 503-05 = 30.5 ft' /

Volume of Ansembly No. 504 82.2 ft' x 0.3 = 24.7 ft' '

Volume of reactor coolant oumo = 112 ft' /

RCS Loop 2A Cold Leg Piping Volume = 248.7 ft' /

g. RCS Looo 2B Cold Leg to Reactor Vessel Pioine Volume NOTE: This includes the piping from steam generator 2 to reactor  ;

coolant pump to safety injection nozzle.

Volume of Assembly No. 503-03 = 28 ft' Volume of Assembly No. 503-07 = 53.5 ft' Volume of Assembly No. 503-05 = 30.5 ft' '

Volume of Assembly No. 504-03 = 82.1 ft' x 0.3 = 24.6 ft' /

Volume of reactor coolant cumo = 112 ft' ./ 1 RCS Loop 2B Cold Leg Piping Volume = 248.6 ft'./

h. Pressurizer Surge Line Pining Volume Volume of Assembly No. 505-12 = 4.4 ft'
i. Steam Generator Volume (from Reference 3.5) l Hot Leg Nozzle Volume =

25.8 ft' 2 Cold Leg Nozzles Volume = 20 ft' #/

Primary Head Volume = 446 ft' 2 Manways Volume = 2.6 ft' /

Thus, the estimated water volume in each steam generator is:

/ / / / i 25.8 ft' + 20 ft' + 2.6 ft' + 446 a' = 494.4 ft' # l l

Since there are 2 steam generators, the total volume = 988.8 ft' ' 4 l

1 1

Page 9 of13 ,

6/27/96

O Calculation # %.ENG 1392-M2, Rev. 2 Prepared By: $b6 khh Reviewed By: D7/ gh

j. Total"Stannant" Water Volume in RCS 2,370 ft' /

, . . S,um of" stagnant" RCS volumes = ,

Conversion factor is 7.48 gallons /ft'/

Estimated gallons of" stagnant" water in RCS = 17,731 gaHons /

6.5 Volume ofBoric Acid Recuired to Increase SDC to 2100 nom Ci.is.: = 1450 ppm (measured on June 3,1996) l Cam = 5500 ppm /

Cr=1 = 2100 ppm , '

Vsoe = 34,623 gallons / - I' Vsm x 5500 + 34,623 x 1450 = (Vam + 34,623) x 2100 l

l 5500 Vsm + 50203350 = 2100 Vsm + 72708300 3400 Vsm = 22504950 ,

l l

Vsm = 6619 gallons 6,619 gallons of boric acid at 5500 ppm are required to increase the SDC '

boron concentration from 1450 ppm to 2100 ppm. /

6.6 Evaluation of Reauired SDC Boron Concentration NOTE: This is an evaluation to determine the minimum boron concentration required to ensure that the resulting SDC boron concentration remains greater than the refueling concentration of 1730 ppm (Reference 3.7).

Ci.ac.i = 7 ppm e

C.yoi = 1300 ppm /

Cr i= 1730 ppm /

Vsoe = 34,623 gallons /

V. . = 17,731 gallons /

Vr.i = Vsoc + V.p. = 34,623 + 17,731 = 52,354 gallons /

Vr.i x Cr " Vsoc x C;.;o.i + Vmpoi x C,o, i V<.i x Cn i - V x C-V oc Page 10 of 13 6/27/96

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

9 I

Calculation # 96-ENG-1392-M2, Rev. 2 l Prepared By: g4 (/27/94 {

Reviewed By: yyy g/w/p6 Cm =-. 52,354 x 1730-17,731 x 1300 ,

\

34,623  ;

i

~

C;,is.: = i950 ppm 'N

The SDC/RCS boron concentration needs to be maintained greater than 1950 ppm in order to satisfy the refueling boron concentration l requirements of Technical Specification 3.9.1.  :

]

Page !! of13 4 6/27/96

i Calculation # 96 ENG-1392-M2, Rev. 2 Prepared By: p 6/2?f[%

Reviewed By: F# 4/28 fe l

7. REVEWER'S COMMENTS AND RESOLUTION:

COMMENTS RESOLUTION W g/gerdi& - ChnLk-

k artfs N CoAr hksig. 1&

drwh~ fob lo 3M&

' k lo AlA*w w M/ed evus/ beres. Ccuc tac M onsto/

l l

i l

l l

l i

I Pacsilof13 6/27/96

1

\

i Calculation # 96-ENG-1392-M2, Rev. 2 l Prepared By: N '

Reviewed 77k By: % #4

8. ATTACHMENTS:

8.1 Combustion Engineering memo F-MD-94/N-MD-146, dated June 5,1972,

" Water Volumes and Flow Areas for Hutchinson Island and Millstone

, - Unit 2 (02-32-00)" (CDCC # 43173) 8.2 Millstone Unit 2 Reactor Coolant Piping Instruction Manual, TabTe 1.1 and Drawing E-234-000, Rev. 4, " General Arrangement ofPrimary Piping" 8.3 Babcock and Wilcox Performance Data Report, BWC-222-7612-PR-1, Figure 3.6 8.4 Millstone Unit 2 Final Safety Analysis Report, Table 4.3-4 i

i l

I Page 13 of 13 6/27/96

{

D ,

..%-Ere-13t2-m2,b. 2  % ?.1 Re INTEROFFICE CORRESPON. i.9 i M

l

=2 COMBUSTl0N OlVISION . nn

.bblu.m,u_.

G, 3. FaderCbCC%IOI&Wa

- er, Volumes and Flow. Areas . L. MIavrence '

l To: D. 8. Hood for Hutchinson Island and' .~

, E. B. Mulliken Millatcoe Unit 2 June 5, 1972 --

i ,

(02-32-00)- .

~

i zer 31stributien . F-ND-9ts "'~

E-MD-1k6 R:ference: Memo (1): (H-LFE-005/IE-LFE-119) Hutchinton Island and Millstone Unit 2

FSARS, fran G. B. Fader, D. S. Hood and H. B. Mulliken to' O. 3. Brinknan and R. Kaamann,i July 26, 1971.

Memo (2): (F-MD-71/N-MD-110) Water Volume s and Flow Areas for Hutchinson Island and Millstone Unit 2,, frein J. L. DeLawrence to C. B. Briniana and R. W. F===nn, September 1,,1971.

d. l As you requested in Reference Memo (1), we have calcu34ted FSAR data applicable to both Hutchinson Island Unit 1 and Millstone Unit 2. Prelizzinary values for the requested dats, per Reference -(1), was transmitted per Reference (2). Enclosed are the final data-(water volumes and flow areas with CEA inserted), resulting frcan check Calculations 16-40-13 and -12, based en n=4 e=1 (design) dimensionn specified in the c-E drawings and Chattanooga vessel drawings. Also, enclosed are five (5) figures and a block diagram l fer clarification of data.

This memo (F-MD-94/N-MD-146) supersedes Reference Memo (2) and all asterisks indicate th3 changes made frcen Reference (2).

JLD/cib ,

ottachments T;"

y f

$ 5 s

e

%- ENG-1392-r*2, ges. 2 A&. P. l P.

J 1.6 w FIDr AREA StDNARY See Figure I i. 2 Area in ft.

1. Inlet nozzle area in 19 6
2. Inlet nozzle area qut. ,

27 0-

~

3 Area at nozzle annulus .-

  • 26.89

'4. Fuel. alignment at annulus

  • 3'4 97 4a. Isakage' around outlet nozzle . .225.

i ,

3a. Leakage,past alignment pins. .061 CSB and Thermal abield annulus 8.82 5

6. Thermal shield 'and Vessel annulus *'20 30

~

7 Annulus Cs plate elevation. *.39 58

8. Thru and around flow skirt
  • 46.23 109 74 9 Bottom head
10. Ares thru lower support plate 28.49 Core Support' Plate 28.30 11.

11a'.' Cross Section at colume support #101.08

12. M akage past core support plate .067
  • 3 98 13.' Annulus esa and lower support cylinder 1,11 13a. Core shroud bypass gap.

13b. Max 4nnnn core shroud bypass gap

  • 14.43 53 515 14 Active core
  • .064 14a. Leakage fr:an active core to core shroud bypass -1 Flow into hot plenum thru fuel align. plate and between 23 04 2 15 C plate and CSB.
  • 1.608 2 15a. Imakage.between fuel plate"and CSB

.: }

  • 8.018  :=

15b. Area between core shroud and fuel alignment plate s 16 Core support barrel outlet nozzle 25.64 (

17. Upper guide structure hot plenum at nozzle elevation
  • 79 589 j i

@[p-G&lW1.*M2 $U. I- o N. S. I 93 1 5 so 18.. Flow into single thru fuel alignment plate

  • 3 366

.19. cFlow into dual thru fuel alignment plate  ;.*

1.3s17

. 20. Inside of single at nozzle elevation

  • - 19 571
21. Inside of dual at. nozzle elevatica # 11.'95 W
22. Leakage fr:ma head to UG8 het plantaa - Neu. - ,
  • 1.042 -

Flo. - -

  • 935  :,

_.,; _m 23 Flow thru instrumentation plate

  • 16.800
24. Vessel' outlet nozzle ,

19 242 22s. Frean CEA single to head

  • 22 383 O

e 9

I e  !

l i

1 +

I 4' c i .

3 1 e P

2 i

. = 1 f

f C- 1 1

3 '

! I

_ _E l

4 L _ ..- _

' UEt46 -IT)2-mt.,9m.2  % r. ,

h'"', F3 4 4 to f Tom, wAm vou3es i' volume in ft.3 ._ .;, . --

70 Inlet nozzle

  • Nozzle to fuel alignment plate elevation (annulus) 29 , *. 133 33

Fuel alignment plate to active core (annulus) 12. 50 100-Therent shield and CSB (annulus)'

' 241 Thermal shield and vessel (annulus) 24 Bottcza of active core to core plate (annulus) ,

198 Core support Plate to flow skirt base 333 Inside flow skirt - - -

194 Head below flow skirt 317 Cors support structure 178 Coro bypass shroud I 3i ' '

Active core bott:sn *

  • 644 .

Active core 82 Active core top

  • 362 17pper guide structure to centerline of nozzle -
  • 442 e Upper guide structure from nozzle centerline to suyport' plate
  • 9'T .

Outlet nozzle .:.

  • 64' Single CEA shroud to nozzle centerline
  • 376 Dual CEA shroud to nozzle centerline Sin 81 e CEA shroud 'above centerline of nozzle
  • 181*

3n

~

  • 92' Dual CEA shroud above centerline of nozzle 2 -
  • 203 *  ;

Head below instrumentation plate 3~

Head above instrumentation plate (CEDMs are included) 425 c' E.

4,667

96-Er4&-f3'l2-m2, Rev. 2. Aff. p.[

S *f '*

O O f

. . 3 COMBUSTION ENGINEERING. INC.

WINDSOR. CONN.

CUSTOMEM CONT.No. _MAeg Bf -- Darg ACATION DWG. No., C w o BY CAtt

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k WENCr-I342.-m2, Kv. 2. 44. f. I P s t. to J

\ 3

. Elevation Region _

Water Volumes

~

Volume in Ft.3 1, 114 l Elev. 1 Elev. 2 e 1,14'O a

. uu, mte 3

- .: 671 E1.+. u .

  • 1,296 )

Eley. 5 l

l

= 4,670 ft'.3 ,

l TOTAL VOIDME Therms 1 Begions ,

Wathr Volumes -

Volume in Ft.3 '

I

+ 1,873 mer. 1

- 644 Ther. 2 '

167 Ther. 3

  • 81 Ther. 4
  • 1,206 l Ther. 5 . "
  • 673 l Ther. 6
  • 26 Ther. 7

=

4,670 ft.3 ,

TOTAL VOLUME  :

. l i

1 i

.m.~,.. -- -1 _ ~ _ _ .._ _. _ -m.

_ _ . .m

.  % ErKr-I392-M2, Gv. 2 -

R H., f. l 94 41o

. .- . m - n ,c ,c n e c, .

. . . . TP ,._ __ l v CEDM N0ZZLE .-

gssp:;'

lNSTRUMENTATION" H.

N0ZZLE

.y %F

' I

\ In - s a s i n . p u g s .

)l[h

~

.'. CONTROL .

AllGNMENT -

1 PIN ELEMENT- "Im iM ES .

'- *j;[ UPPER l

ASSEMBLY FULLY 'l 4a I

[P "

i:

!i I

GUIDE

~

WITHDRAWN d 4 id)i 2-3 STRUCTURE 9 y' 'i.'

j Q.Q i;'

.'..+, m . - - - -

. .f_ .

' -4 e .  !! wnj .

A

  • C ot pd,g.yU , F@/

OUTLET .lNLET a

'f N0ZZLE' g --. s.

s N0ZZLE

% .. c mpa=.a.:a0my0 l% ,

w~

I J CORE

[l* LIM)_[ 1.Ef, .

M hkkfL -

lr $).' I~IQI

.., . . f!I' .1s =

THER. SHIELD.

'/

ACTIVE

.2 o) 1 rl T% -

CORE

%c -l:N

e
.

., , l -SH.ROUD CORE

.i '. .

,=.

' LENGTH I ii Vi ** N 2 -

i l, ! l FUEL Ea f i

f. .'

= mrg) i

]"'9"/2 ASSEMBLY 4

.. ..  ;;i- .

ht

~

I' l -

SNUBBER ORT -

\f: i:i!!!!!!!:  %':k'_).g::.:i!!!!!!) 8

' ASSEMBLY

)

l

! CORE STOP 7 .

FLOW -

/

/ .,-

SKIRT

~ ' -'

FLOW AREAS

. i m COMBUSTION l figure '

h swa;N5ERING I Reactor Arrangement I

' ' Nuclear Power Depart.m e n t I

.. l %.-EN G-1392-p2,. A '2-

, .. i. . - A+.9.1

. .... ... .- . P t el.t.o 3-

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[,. .g. i e. .. ,,:, . . , . .,,

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n-,nnnn-r w-4p g... W- -CEDM N0ZZLE

/ . -

55: $,@s "

.. INSTRUMENTATION

.~-

INSml.flA K . . --

1

%f N0ZZLE -

w

-. w. .

CONTROL . . '

AllGNMENT "-

~ '-

. / .. V's

. ASSEMBLY ELEMENT "t mil i..  !

f- . PIN '~7 Q~[M t 4

}-, ",3  ! -UPPER -

!-r. FULLY h!~ " '

%l! lj i GUIDE

...? '.

WITHDRAWN i j u e. .

. STRUCTURE. . -

. ... .: ...v. -

)y p:.

l f-

~ .

v- - -

li ,

.i .. . ;-

. : ~r;:  %. -

- l '

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>y n e >-  : .'

OUTLET '. ' + ' ' ' -

INLEf

/

.. ?, ..

'% N0ZZLE -

_ "-. m _ L A=*..;,__ . .

N0ZZil r -

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. g.,

y .'. .

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  • s' . .. J w
3 ..

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l i

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, j. , .AhlVE gE  ;{,

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i c! i i .I i MN

ij  : l- TI FUEL

. f.  ;;  : ili ASSEMBLY l

1. . ,

~~ytyWif%f  :

, q -1..i..i%

3 T .N SUPPORT A

CORE d-Cjd SNUBBER I. 4 "

q e .- y, " ,,s. .- -

\ ['.- .

ASSEMBLY-CORE STOP

.!!!!!!!!!!!!!!!!!!!!!!!!!!!!s N. .

FLOW -

SKIRT .

~ & C. 3' '

__.ELiivKn0N _ REG 10NS.'?Z.

tm1 CO M 7 Reactor Arrangement e.; 2xa n.'I.!STION .s s,ar.y a l iFigu r-

E l  : nuclea r Po.ve r o eoar t.ienti Illustration is typical of design, but does not necessarily show exact construction of equipment'.

.EAG-I'.n2- m2, @sa. 2. . A+h 3. l

& 3" 9, loc

, /. ).  :.

4 t.  ; .. .

'b .; ,- .- -

ZTIARQCET1f 1T"l

S< . .;

r7 9 J._ g e CEDM N0ZZLE n !NSTRUMENTAT10i o

.. . ; IFmiN R.i. ATE.'..".- -

4,.

,y.p$g$Q; , - /

N0ZZL.E ..:, -

ge

/'c i.

W/

.. c .

q' uf.(:,j.v! [2.rb ALIGNMENT- '

CONTROL. .. ~ .. . PIN l ELEMENT _ W - 'n % ~} n YpT~2  :- i n /N IM  !

've ASSEMBLY u i' is -UPPER

~

.. . FULLY- N Q$": \ ! N -

~

.. GUIDE '

c,...

- . W!.THDRAW&~ b

..:n J  % Ki n }:N \L . STRUCTURE

.,.r..

N y< ,J K, y%*., ' . n- -: op .

Q g .. .

.. c- ..

. .. : N.

s. h , .

.g :

- [ys 5

. .~..j . .,

. y_ ; . .- .

( l %'E jjjj>d.q' Y 5.' / 'lNLEY

. 00TLET' N0ZZLE'

..- e

. N0ZZLE '

- q$

j'9!@i(9 K V4 Fd.<.I CORE i

.g. *:

". 9 ;3- : : -

1 I -NM7Iy( x ' _N_ .:.;

SUPPORT BARREL ,. ; -

- .0

..b

3. . .h';h 'l.s [E '

M L + - T W e Ri M . S m E .t.0. -.

4 J. I s A.l1__

ii W Wn CORE-

. .~ >

. ACTIVE  ;. .-

-SHROUD L

e CORE -

- -- - - _i .

A ifil pq

'P -

. LENGTH =._

4+ '

II FUEL -

i rb '-

ASSEMBLY.

d % l~.TT NIi f

[

. .. . .. . f. .

,.,. , Q. . '"

. g ;.;,; ...r ~~

.- .. .,j R. r.i'~i- .

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ASSEMBLY i

.. .. .z

.. . 4:_::_!!!!!!!_; 'i:!!!!!!:::::q .

. FLOW CORE STOP ~P -- ' -

... SKIRT ...

T-

..  : .'J...U." THERMNL REG 10N -

'-. 1. .

Figur 2

- p cc:veuSTION Reactor Arrangement

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c -

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y- . Anlet nozzle . , fuel al'ignment plate .

f.. . .

lower end fitting res

, upper cura support baltral ' ,

annulua re6 1,on .

l'over core support be

, , -core shroud

. . CSB. flange- . -' r center core , ' flow skia support barrel extension guide- . r-lower yea ed .

-3 instrumentation nopzle \ .

I I l I w - -

a  ;

\ 1 ) p -

G -- -

l e s = = =-+ = = -=~'

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) -

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l 1

ii..

FIGURE 3.6 PRIMARY SIDE WATER VOi,UME l PRIMARY SIDE PRIMARY SIDE VOLUME WITH VOLUPEINCLp)ING N0ZZLES (FT N0ZZLE DAMS

  • IN PLACE (FT ) l Primary Inlet Nozzle 25.8 ft / nozzle x 1 nozzle 26 -

Primary Outlet Nozzles 3

10.0 ft / nozzle x 2 nozzles 20 -

Primary Manways 3 3 1.3 ft /manway x 2 manways, 3 Primary Head (minus divider plate 446 446 and stay cylinder volume)

Tube Volume Tube Volume /Ft. gerStraight

' Leg': 20.02 ft /ft Total Tube Volume in Straight 1038 1038

' Legs' Tube Volume in U-bend 160 160 TOTAL: 1693 1647

em t

%Er4G-13%mt, fa 2 M. P. 4 MNPS 2 FSAR Fa.p / .Il TABLE 4.3-4 REACTOR COOLANT PUMP PARAMEi t RS g.

~ Number 4

Type '

Vertical, Limited Leakage, Centrifugal Shaft Seals . -

Stationary Face ,

Mechanical (4) '

Rotating Face Body Carbon CCP-72 Rotating Face Ring . ASTM A-351 Gr CF8 '

Titanium Carbide, 4 Kenna-Metal K162-B Design Pressure, psig 2,485 Design Temperature, 'F '

650 Normal Operating Pressure, psig 1 2,235 Normal Operating Temperature, 'F 549 Design Flow, gpm lQ 81,200 Total Dynamic Head, ft 243 Maximum Flow (one pump Operation), gpm 120,000 Dry Weight, Ib 168,050 Flooded Weight,Ib As-M 175,050 Reactor Coolant Volume, ft3per pump 112 Shaft Material ASTM A-182 Type F-304

, Material ASTM A-351 Gr CF8M C tsing Wear Ring Material ASTM A-351 Gr CF8 Hydrostatic Bearing Bearing Material Journal Material ASTM A-351 Gr CF8 ASTM A-351 Gr CF8

.M

.,

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