ML20154C102
ML20154C102 | |
Person / Time | |
---|---|
Site: | Oconee ![]() |
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)
P PDR
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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
- 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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