ML20154C102

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Proposed Tech Specs Changing UFSAR by Increasing Max Rod Internal Pressure in Spent Fuel Pool from 1,200 to 1,300 Psig
ML20154C102
Person / Time
Site: Oconee  Duke Energy icon.png
Issue date: 09/30/1998
From:
DUKE POWER CO.
To:
Shared Package
ML15112A259 List:
References
NUDOCS 9810060145
Download: ML20154C102 (3)


Text

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Oconce Nuclear Station v

15.1I Fuel Handling Accidents 5

Where:

t h bubble rise time, seconds i,

5' d = effective bubble diameter, em 5

Since the mimmum water depth over a dropped fuel assembly is less than 23 feet (21.34 feet), the assumed 5

iodine DF must be less than 100, according to Reg. Guide 1.25, and calculated with comparable

./

5 conservatism as done in Reg. Guide 1.25. Using the above relationship, with a water depth of 21.34 feet, M*he fuel - "h[ qual to 19 (Reference OSC-6070).

a comparable DF is e 5

J$4et A ff$McM I'N#'

5 T

y. :;r pressufe, at a Spent Fuel Pool bulk temperature of 150 F.

M"M e L 5

less than 1200 psig[ based on the present TACO 2 computer code licensing limit of 2200 psia at operating 5

system condition (Reference FSAR Section 4.2.3.1.3, " Fuel Thermal Analysis").

e85ct i D 5

The hetivity released from the water's surface is released within a two-hour period as a ground release.

5 The atmospheric dilution is calculated using the two-hour ground release dispersion factor of 2.2 x 104 5

sec/m2 5

The totalintegrated dose (2-br EAB) to the whole body at the 1-mile exclusion distance is 0.185 Rem and l.-

5 the thyroid dose at the same distance is 52.45 Rem. These values are far below the limits given in L

5 10CFR100 of 25 Rem whole body and 300 Rem thyroid.

.5 15.11.2.2 Base Case Fuel Handling Accident inside Containment l

l 5

In 1977, the NRC asked Oconee to evaluate the offsite dose consequences for a fuel handling accident l

5 inside containment, per the guidance given in Reg. Guide 1.25. Since the shallow end of the fuel transfer l

5 canalis at an elevation of 816.5 feet, the same iodine decontammation factor used for the Fuel Handling l

5 Accident in the Spent Fuel Poolis used for the Fuel Handling Accident inside Contamment. The activity l

5 released from the refueling water is released as a ground release, which has an atmosphenc dispersion 5

factor of 2.2 x 104 sec/m2. There is no credit taken for any containment closure / integrity resulting in the 5

released activity from the refueling water going straight outside.

l i

5 Using the fuel assembly gap inventory in Table 15-1, and assuming all 208 fuel pins are damaged, the 5

two-hour EAB dose is 0.185 Rem to the whole body and 52.45 Rem to the thyroid. These values are 5

appropriately within the guidelines given in 10CFR100 (appropriately within means 100 Rem to the 5

thyroid), and are identical to the base case Spent Fuel Pool Fuel Handling Accident described in Section l

l 5

15.11.2.1, " Base Case Fuel Handling Accident in Spent Fuel Pool."

l S

1 5.:11.2.3 Supplemental Cases of Fuel Handling Accidents l-5 To provide additional information as to the sensitivity of various input assumptions into the offsite dose 5

consequences of the fuel handling accident, additional supplemental cases are described here.

5 CASE A:

5 If the radioisotope release from the spent fuel pool water's surface is assumed to be captured by the Spent 5

Fuel Pool Ventilation System, resulting in an elevated release, (atmospheric dispersion factor is equal to 5

3.35 x 10-5 sec/m3) and assuming that the Spent Fuel Pool Filters are 90% efficient for the removal of elemental and particulate iodine, and 70% efficient in the removal of organic iodine, the resultant 5

two-hour offsite dose is calculated to be 1.2 Rem thyroid and 0.021 Rem whole body at the exclusion 5

area boundary (EAB).

5 CASEB:

9810060145 980930

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2 of k PDR ADOCK 05000269 (31 DEC 1997)

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l L r, Attcchment.4 l'

Markup of UFSAR Change L'

I l

INSERT A-(Section-15.11.2.1. Dace 15-34)

H Duke will use a DF equal to 89 for a maximum rod internal pressure in the spent fuel pool of 1300 psig-for the fuel

-handling accident analysis per reference 21. lThis was justified in Reference 21 using the WCAP-7828 methodology described above.

P I

INSERT B (Section 15.11.2.1, oace 15-34) cnr is calculated to be less-than 1300 psig (based on the present TACO 3 computer code licensing limit of a proprietary value above nominal system pressure at operating system conditions)~.

t i.

I i

Oconee Units 1, 2,

and 3 3 of 4 i

m_

  • . t
  • o o

15.16 Post-Accident Hydrogen Control Oconee Nuclear Station

-5 15.16.7. REFERENCES 1

1. Shure, K., Fission Product Decay Energy, WAPD-BT-24, December 1961.

1

2. Allen, A. O., The Radiation Chemis:ry of Water and Aqueous Solutions, D von Nostrand Co., Inc.,

1961.

1

3. Morrison, D. L, An Evaluation of the Applicability of Existing Data to the Analytical Description of I

a Nuclear. Reactor Accident, Quarterly Progress Report for April through June,1968, BMI-IB44, July 1

1%8.

1

4. Zittel, II. E.,

Radiolysis Studies, ORNL Nuclear Safety Research and Development Program 1

, Bi-monthly Report for September-October 1967, ORNieTM-2057, Nov. 27,1967.

1

5. Coward, II. F., Jones, G. W., Limits of Flammability of Gases and Vapors, Bureau ofMines Bulletin 1.

503.

I

6. Markstein, G., " Instability Phenomena in Combustion Waves',4th Symposium on Combustion.

~

1

7. Shapiro, A.

M., Mofette, T.

R., Hydrogen Flammability Data and Application to PWR 1

Loss-of-Coolant Accident, WAPD-SC-545, September 1957.

I

8. Coleman, L F., et al, Large-Scale Fission Product Transport Experiments, BNWL 926, pp. 2.1 to 1

2.21, Dec.1968.

I

9. Stinchcombe, R. A., Goldsmith, P., " Removal of lodine from Atmosphere by Condensing Steam",

1 Journal ofNuclear Energy Parts A/B 20, pp. 261 to 275,1966.

1

10. Stinchcombe, R. A., Goldsmith, P., Clean.up of Submicron Particles by Condensing Steam, 1

AERE-M-1213.

11. Goldsmith, P., May, F. G., 'Diffusiophoresis and Thormophoresis in Water Vapor Systems", Aer$ sol I

1 Science, C. N. Davies, Ed., Academic Press, Inc., New York, New York, pp. 163-194 (1966).

I

12. liyland, E. L, ' Design Criteria, Contamment liydrogen Recombiner System (Rev. 0)," Duke Power 1.

Company, June 24,1983.

I

13. II. B. Tucker (Duke) letter to II. R. Denton (NRC) dated October 20,1986.

5

14. Regulatory Guide 1.7 (Rev 2), " Control of Combustible Gas Concentrctions in Containment i

5 Following a Loss-of-Coolant Accident' l

5

15. OSC - 6191 (Rev. 0), ' Reanalysis of Oconee liydrogen Recombiner and Purge System Requirements' S
16.,Wiens, L A. (NRC) letter to J. W. Ilampton (Duke) dated February 7,1996.

5

17. OSC - 123 (Rev.1), " Activity on l'ilter RB liydrogen Purge' 5
18. OSC - 6534 (Rev. 0), 'llydrogen Purge Cart Operator Dose Rate' j

5

19. OSC - 3781 (Rev. 5), " Documentation of Maximum Ilypothetical Accident (MiiA). Dose Model For 5

Oconee Nuclear Station' i

5

20. OSC - 6064 (Rev.1), ' Estimated' Radiation Dose Rates in the Aux 2hary Building Following a Large l

5 Break IDCA" l

1 TIIIS IS TIIE LAST PAGE OF TiiE CIIAPTER 15 TEXT PORTION.

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