ML20005E170

Forwards Data Re Steam Generator Tube Rupture,Main Steam Line Break,Control Room Vol,Summary of Inputs & Documentation of Impact of Using Reduced Iodine Filter Efficiency in Control Room Dose Calculations

ML20005E170
Person / Time
Site:	North Anna
Issue date:	12/22/1989
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
89-022A, 89-22A, NUDOCS 9001040017
Download: ML20005E170 (17)

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VIRGINIA ELECTRIC AND POWER COMPANY RICilMOND. VIRGINIA 23261 h

, December 22, 1989 i

U. S. Nuclear Regulatory Commission Serial No. 89-022A Attn: Document Control Desk N0/PJL Washington, D.C. 20555 Docket Nos. 50-338

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50-339 ,

License Nos. NPF-4 )

NPF-7 Gentleme n:

VIRGINIit ELECTRIC AND POWER COMPANY NORTH AtNA POWER STATION UNITS 1 AND 2 CONTROL ROOM HABITABILITY - SUPPLEMENTAL INFORMATION By letter' Serial Number 89-022, dated March 1,1989, Virginia Electric and Power Company. submitted documentation of actions taken to ensure compliance of the North Anna Power Station with General Design Criterion (GDC) 19 of Appendix A to 10 CFR 50. A reevaluation of control room dose consequences was also submitted, which indicated that radiation exposure to control room ,

operators would be increased above the values currently delineated in Chapter 15 of the North Anna Updated Final Safety Analysis Report. License amendments were' requested to revise the limiting doses to the control room operators to the values in our March 1,1989 submittal.

On November 15, 1989, Messrs. C. Nichols, J. Lee, and H. Walker of the NRC met with Virginia Electric and Power Company personnel to discuss and clarify the bases of those control room habitability calculations. As a result of that meeting, we were requested to provide additional information concerning some of the inputs used for our North Anna control room dose calculations. This information was requested so that independent dose calculations could be

. performed by the NRC as part of its Safety Evaluation Report supporting the proposed license amendment.

The requested data, which are attached, consist of the following:

Steam Generator Tube Rupture hydraulic data used in the control room dose calculations (including the flashing fraction of primary coolant as a function of time) and source terms used, and the basis for our calculated 10 minute tube uncovery time.

Main Steam Line Break hydraulic data used for the control room dose calculations, including break flows.

The control room volume (as opposed to the pressure envelope volume),

for use in calculation of doses due to shine.

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• A copy of the summary of the inputs used for the various dose calculations, which was shown to Mr. Lee during the meeting at North Anna.

Documentation of the impact of using a reduced iodine filter efficiency (95%, versus the 99% used to determine the submitted doses).

During the meeting, Virginia Electric and Power Company also agreed to provide any information which we could locate on the original LOCA analysis regarding the duration of leakage from containment which was used. We have determined that our LOCA dose calculations dating back to 1972 made the same assumptions concerning containment leakage which were used in the most recent calculations: specifically, that the containment leakage rate is 0.1 volume %

per day, and that the duration of this leakage is one hour. We have not located any documentation which indicates whether the NRC made other assumptions about containment leakage when they performed the calculations to support their SER of the original LOCA evaluation.

If you have questions regarding this additional information, please contact us.

Ver,y truly yours,

/A W / Ar -

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W. L. Stewart Senior Vice President - Nuclear Attachments:

1. Supplemental Information: North Anna Control Room Dose Calculation Inputs

2. Summary of Control Room Habitability Inputs for Various Accidents (North Anna Data) cc: U. S. Nuclear Regulatory Commission 101 Marietta Street, N. W.

Suite 2900 Atlanta, GA 30323 Mr. J. L. Caldwell NRC Senior Resident Inspector North Anna Power Station

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, e Attachment I r  :

Supplemental Information:

North Anna Control-Room Dose Calculation Inputs I

1. SteamGeneratorTubeRupture(SGTR) Data:

-a. The initiel total secondary side steam generator mass is 99,100 lb,(91,600 lb m. liquid and 7500 lb, steam).

b. Primary coolant is released to the affected steam generator as follows:

1) 0 to 10 minutes: 44,000 lb,

! 2) 10 to 30 minutes: 88,000 lb, ]

c. The primary to secondary leak rate was taken from the North Anna Technical Specifications: l

1) 500 gpd (or 0.35 gpm) in the affected steam generator

2) 1.0 gpm total

d. The- following steaming masses were used for the SGTR dose calculations:

Steaming Mass Time Span Faulted SG Intact SG (hours)

(lb )

_____m_.... ___

(lb.m____

)

0-2 81,640 341,000 2-8 0 605,500 These values were calculated assuming the faulted steam generator is isolated in 30 minutes, and that the intact steam generators cool the RCS from 30 minutes to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The RCS is cooled from Hot Full Power (HFP) to Hot Zero Power (HZP) in the first two hours after the accident, and from HZP to 350'F T avg between 2 and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />,

e. The coolant activities used were based on the North Anna Technical Specification limits (see Table 1). The pre-accident iodine l- spike calculation assumed an increase in the primary coolant l activity to 60 pCi/g I-131 dose equivalent. The iodine appearance rates used for the concurrent iodine spike calculation are given in Table 2.

f. The addition of downcomer baffle plates to the North Anna Units 1 and 2 steam generators decreased the initial water mass inventory at Hot Full Power, with a resultant period of tube uncovery (water level below the top of the tubes) following a tube rupture event. A 10 minute tube uncovery time was conservatively

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-e calculated for North Anna, and was und in both our 1987 SGTR offsite dose calculation (submitted to the NRC in our letter serial number 87-474C, dated September 25,1987) and the control i room dose calculation submitted in 1989. l

. To . determine the period of tube uncovery, the faulted steam

. generator secondary inventory .was calculated as a function of time from the trip. This calculation considered the following contributions:

1) the initial mass (99,100 lb,);

2) the mass dissipated to atmosphere to cool the unit to HZP conditions, based on the energy removal required to reduce the core stored. energy, the RCS metal stored energy, and the RCS fluid energy;

3) primary to secondary break flow, based on 132,000 lb, total  ;

break flow;

4) auxiliary feed flow to the faulted generator, beginning 60 seconds after the trip, and terminating only after the narrow range level is recovered in the faulted steam generator; and i

5) boiloff of secondary fluid to remove decay heat, which was calculated as a function of time.  !

This inventory was then compared with the calculated mass required to keep the tubes covered to determine the uncovery-time. 3 The predicted tube uncovery time for the full break (which was '

determined to be the limiting break size) was about 9 minutes.

For the dose calculations, this value was conservatively increased to 10 minutes.

2. Main Steam Line Break (MSLB) Data:

a. The primary coolant volume is 9380 ft 3,

b. The fe flow rate to - the affected steam generator is 9.35x10gdwater lb,/hr.

c. The feedwater isolation time for the affected steam generator is 9.3 seconds.

d. The primary to secondary leak rate was taken from the North Anna Technical Specifications:

1) 500 gpd in the affected steam generator

2) 1.0 gpm total (3 steam generators) = 1440 gpd

e. The steam generator masses used were:

1) Liquid = 1.67x10 5lb / msteam generator

2) Steam = 4238.6 lb /m steam generator

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b; These values were taken from the North Anna UFSAR.

f. The steam generator (secondary side) volume is 101 m3 ,

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g. lhe steam release from the affected. steam generator is 358,000 lb, during. the _ first 30 minutes (until the affected steam generator is isolated), and 0- lbm for the remainder of the analysis. ,

, i l h. The energy and mass release rates as a function of time are:

Time Mass Release Energy Release (sec) (ib/sec) (BTV/sec) 0 10511 12.568 x 106 36 1079 1.E15 x 106 90 822.2 0.9229 x 106 i 180 195.4 0.2129 x 106 236 123.7 0.1336 x 106 i 521 123.6 0.1338 x 106 1311 123.5 0.1346 x 106 1701- 123.4 0.1349 x 106 1800 123.4 0.1350 x 106 1

Thgseratesarebasedonoperationat102%fullpower,anda1.4 ft double ended rupture.

i. The auxiliary feedwater flow to the unaffected steam generators I is 900 gpm for the full 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> releases occur from these steam generators.

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J. The coolant activities used were based on the North Anna Technical Specification limits (see Table 1). The pre-accident iodine spike calculation assumed . an' increase in the primary . coolant activity to 60 pCi/g I-131 dose equivalent. The iodine appearance rates for the concurrent iodine spike calculation are given in Table 2.

k. The turbine building volume is 4x106 ft , 3

3. The cont room volume for calcWadon of dose due to sMne is 1.165x10{ol ft. 3This volume represents only the control room, and does not include the entire pressure envelope.

4. A copy of a table summarizing some of the inputs for various North -

Anna control room dose calculations is given in Attachment 2. A 4 version of this table was shown to Mr. Jay Lee of the NRC during our November 15 meeting at North Anna, i

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5. The impact on control room doses when the assumed iodine filter r

efficiencies are reduced from 99 to 95 percent can be seen in-Table

'i 3. As noted during the November 15 meeting, dose calculations are in progress to support a recirculation rate lower than the 2000 cfm used to calculate the doses submitted. earlier this year. A lower iodine filter efficiency is being used in those calculations to more accurately reflect the North Anna- charcoal filter design . and be

consistent with Regulatory Guide 1.52.

v i The sensitivity of of the 30-day control room doses following a steam generator tube rupture to variations in the. control room recirculation rate and charcoal filter efficiency is illustrated in Table 3. The doses in the first column are based on operation of two fan / filters in the recirculation mode (1000 cfm each) with an iodine

. filter. ef ficiency of .99%. This is the second steam generator tube rupture case described in our March 1,1989, submittal, which assumes automatic initiation of recirculation upon receipt of an SI signal.

This assumption incorporates some planned modifications of the North .

Anna control room ventilation system which were recommended as a I result of the control room habitability evaluations. The second l column in Table 3 similarly contains doses based on operation of two fan / filters (2000 cfm), but assuming a 95% filter efficiency. The doses in the third column were calculated assuming only one fan / filter operating (1000 cfm) with a 95% filter efficiency for iodine. This third case is representative of the calculations currently in progress. All doses listed are for the pre-accident iodine ' spike 1 case, with control room recirculation starting automatically' upon receipt of an SI signal 6.5 minutes after the tube rupture occurs.

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-Tablet Coolant Activities Based on-f; , ' -North-Anna Technical:-Specification Limits' --

c k Primary Coolant Secondary Liquid Secondary Steam Concentration Concentration Concentration-Nuclide- (pCi/g)~ (~Ci/g) p (vCi/g).

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Kr-85m! -0.564 -

3.72 E-4 Kr-85 ~1.37. --

9.04 E-4.

[ Kr-87 0.326 -

2.15 E-4 h Kr-88 .0.986 -

6.51 E-4.

.'. Xe-131m- 0.837 -

.5.52 E-4

-Xe-133. 75.'5 -

4.98'E-2 Xe-135m 0.0507- -

3.35 E-5

Xe-135 1.64 -

1.08 E-3 Xe-138' O.180 -

1.19'E I-131 0,656: 7.08 E-2 7.08 E-4 I-132 0 239- 8,60 E-3 8.60 E r I-133 1.06 9.53 E-2 9.53 E-4 1

'I-134 0.148- '2.54 E-3 2.54 E-5 I-135 0.571 3.67 E-2 3.67 E-4

, Table 2 Concurrent Iodine Spike Appearance Rates Appearance Rate i

Nuclide (Ci/sec)

I-131 1.16 i I-132 2.29 I-133 2.70 I-134- 3.30 l 1-135 2.53 1

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Steam Generator Tube Rupture l

30-Day Control' Room Doses g' O t e:.

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' Rscire. Rate (cfm): , 2000 '.2000' 1000: "

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_ Filter Efficiency (%) 99 -95 95: _.

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d# 1 Thyroid _ dose;(rem):,, , 16.5-

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4 rf , Gamma dose'(rem)' 'O.012 0.012 ' 0. 012.- :1

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Betadose(rem) - 0. 57_ 0.57 -0. 57 - i 1

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Attachment 2 !e

'k Summary of Control Room Habitability. Jp

.. 1 Inputs for Various Accidents  ;

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. s l: io I. COMMON PARAMETERS FOR ALL CALCULATIONS

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1. Core Thermal Power Level (MWt) 2958 (includes an additional 2% for instrument error)

2. Control Room Unfiltered In-leakage 10 f

(cfe) 0-720 hr

3. Control Room Volume 1.165 x 106

. 4. Control Room Emergency Vent 11a- 99*

tion Filter Efficiency (%):

p Elemental Iodine

• Except LOCA - 90% used.

5. Breathing Rate (m3 /sec) 3.47 x 10**

6. Duration of Bottled Air (br) 1

7. Control Room Recirculation Rate 2,000 (cfm)

8. Control Room Emergency Ventilation 1,000 Rate (cfm)

9. Occupancy Factors 0-24 brs 1.0 24-96 brs 0.6 96-720 brs 0.4 (included in X/Qs)

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i II. LOSS OF COOLANT ACCIDENT (LOCA)  !

1. Containment Volume (ft3 ) 1.84 x 106 h

2. Containment Leak Rate (Vol/ day)

O to I hr .001- '

I to 720 hr 0.0'

3. Composition of Iodine Release to l Containment:  ;

Elemental 91 Methyl 4 Particulate 5 c' >

4 Traction of Core Activity Available  !

j for Leakage from Containment (%)

• Noble Cases 100 Halogens 25*

• with 50% plateout. '

5. Containment Spray Removal Coefficients (br~1)

. Elemental Iodine 10-Methyl Iodine 0

• Particulate Iodine 0

6. Containment Mixing Rate 2 (Unsprayed Volumes /Hr) '

7. Containment' Spray Coverage (%) 70

8. Containment Sump' Volume (liters) 1.5 x 106

9. Sump Water Temperature (*F) <212

10. Traction of Core Iodine in Sump 50

(%)

11. Traction of Iodine in ECCS 10 Leakage Released to the Atmosphere (%)

12. ECCS Leak Rate (cc/br) 0-10 min.

2 x Tech. Spec. 50 gpm 10 min 30 days 900cc/br

13. Safeguards Area Exhaust, 90

• Iodine Filter Efficiency (%)

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LOCA (Cont) 14 Control Room Emergency *v'entil-ation Filter Efficiency (%)

Elemental Iodine 90 '

Methyl Iodine 0 Particulate Iodine 99  ;

15. Atmospheric Dispersion Co- ,

efficients, X/Q values (sec/m 3)

a. Containment Leakage 0-8 br 7.97 x 10'3 8-24 hr 6.30 x 10'3 24-96 hr 2.47 x 10'3 96-720 hr 7.17 x 10-4

b. ECCS Leakage ,

0-8 hr 7.88 x 10"* '

8-24 hr 5.75 x 10-*

24-96 hr 2.05 x 10-4 96-720 hr 3,15 x 10-5 i

16. Control Room Isolation Occurs 0  ;

at Time

17. Bottled Air Actuation Time (hrs) 0

18. Recirculation Start Time (hrs) 0 [

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! III. MAIN STEAM LINE BREAK (MSLB) ,

1. Primary to Secondary Leak Rate 500 via the Affected Steam Generator (gpd) '

2. Primary to Secondary Leak Rate 940 t via the 2 Non-affected Steam Generators (gpd) j l

3. Duration of Release.via the 30 Affected Steam Generator (min.)

~ 4. Duration of Release via the 2 .

Non-affected Steam Generators (br) 8 1 5. Off-site Power Lost *

• Therefore, the condenser is not available for cooldown.

6. Pre-accident Iodine Spike 1131 60 Dose Equivalent (pCi/cc) '

7. Duration of Concurrent Iodine 4 Spike (br)

8. Break Location TurbineBkdg.

9. X/Q values (sec/m3 )

Main Steam Reliefs 0e8 br 7.97 x 10'3 8-24 hr 6.30 x 10-3 24-96 br 2.47 x 10-3 96-720 hr 7.17 x 10'4

10. Steam Condition Pressure (psia) 850 Temperature (*F) 525

11. Control Room Isolation Time (hrs) 0

12. Concurrent Iodine Spike l- Appearance Rates (Ci/Sec) l 1131

' 1.16 1132 2.29 I133 2.70 '

1134 3.30 I135 2.53 1

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13. Iodine Partition Factors i

Covered 0.01

Uncovered 1.0

14. Secondary Side Mass (lbs/SG)

Liquid 167,000  ;

Steam 4238.6 i

15. Primary Coolant Volume (ft3) 9380

(@ 577'F)

16. Total Steam Flow Rate (1b/br) -

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17. Volume of Liquid-in the Secondary Side of SG (m3 ) 101 .

18. Auxiliary FW Flow to Non  :

Affected SG's (gpm) 900

19. -Steam. Release from Affected SG (1bs) 358,000 0-05 hr

20. Aux FW Flow to Affected SG (lb/hr) 9.35 x IOS ,

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I IV. TUEL HANDLING ACCIDENT (FRA) l

1. No. of Damaged Fuel Assemblies 1

2. Radial Peaking Factor 1.65

3.  % of Total Activity in Gap Noble Gases 10 I- >

Kr-85 30 i

Iodine 10 l-

4. Fuel Pool DF Noble Gases 1 Iodine 100

5. ' Constituents of Iodine Above the-Pool (%):

Elemental 75 Methyl 25

6. Time after Shutdown Accident 100 Occurs (hr)

7. Time after Accident the Bottled 10 Air is Actuated (min.)

8. Duration of Release from Fuel 2 Building (hr)

9. Fuel Building Exhaust Filter Efficiency (%):

Elemental Iodine 90 Methyl Iodine 70

10. Fuel Building Exhaust Rate (cfm) 29,000

11. Fuel Building Volume (ft 3) 1.85 x 105

12. Control Room Isolation Time (br) 0

13. X/Q values (sec/m3 )

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' 0-8 br 7.97 x 10'3 l

8-24 hr 6.30 x 10-3 24-96 hr 2.47 x 10'3 96-720 hr 7.17 x 10"

14. Recirculation Start Time (min) 10 l

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V. STEAM GENERATION TUBE RLTTURE (SGTR)

1. Tubes Uncovered @ $ min

2. Duration Tubes Uncovered 10 min

3. Control Room Isolation @ 6.5 min

4. Bottled Air Actuation @ 6.5 min

5. Control Room Recire Rate 2000 cfm ll Control Room Recirc Actuation @ 16.5 min

6. Emergency Ventilation Filter Eff. 99%

7. Control Room Normal Ventilation Rate 1460 cfm

8. Off-site Power Lost

9. Duration of Release via the Affected Steam Generator (min) 30

10. Duration of Release via the 2 Non-Affected Steam Generators (hrs) 8

11. Steam Release from Affected Steam Generator (1bs) 0-0.5 hrs 81,640

12. Steam Release from 2 Non-Affected Steam Generators 0-2 hrs 341,000 2-8 brs 605,500

13. Pre Accident Iodine Spike 1131 l

Dose Equivalent (uCi/cc) 60

14. Duration of Concurrent Iodine Spike (hrs) 4 l 15. Concurrent Iodine Spike l Appearance Rates (Ci/Sec)

I131 1.16 1132 2.29 1133 2.70 l 1134 3.30 i 1135 2.53 l

l 16. Primary to Secondary Leak Rate Via the Af fected Steam Generator (gpd) 500 L

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l -'- Generators (gpd)' 940 k

L -, ,. 18. - X/Q Values (sec/m3 )

Main Steam Reliefs .

i 0-8 hr 7.97 x 10'3'  !

'8-24 hr 6.30'x 10-3-- l s

24-96 hr 2.47 x 10-3 i j- c; '96-720 hr T.17 x 10* *

19. . . Iodine Partition Factors

' Covered 0.01 l F Uncovered .1.0 i

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