ML20101R851
| ML20101R851 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 07/08/1992 |
| From: | Tuckman M DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9207160238 | |
| Download: ML20101R851 (5) | |
Text
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Dahe hneer Cornpuny M S Tecwn Catawba Nucitar G,wrction Department Vice President i
- HiNl'orajordfa>ad (M3hul 3205 Offace 1brk.SC ?3745 (803) UI.3t:6 h e DUKEPOWER l
July 8,1992-U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.
20555
Subject:
Catawba Nuclear Station
- Docket Nos. 50-413 and 50-414
~ Supplement to Technical Specification Amendment Request Unit 1 Cycle 7 Reload Per discussions with your sta'f, additional information is being provided regarding the following items:
1)
Fully withdrawn rod position varying between 222 steps withdrawn and 230 steps withdrawn, and
- 2) - The assumption that 50% of fuel fails due to a rod ejection accident.
If there are any questions regarding the items above, contact Mary Hazeltine at (803)831-
- 3080, Pursuant to 10 CFR 50.91(b)(1) the appropriate South Carolina State official is being provided a copy of this amendment request.
Very truly yours,
// O b & a ar4&
- u. S. ruckman Attachments MHH/CIC7.SUP f()/
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9207160238 920700 PDR ADOCK 05000413 l
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-Q U. S. Nuclear Regdatory Commission July 8,1992 Page 2 xc: Mr. S. D. Ebneter Regional Administrator U. S. Nuclear Regulatory Commission Region II
' 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. Heyward Shealy, Chief Bureau of Radiological Health South Carolina Department of Health &
Environmental Control 2600 Bull Street Columbia, South Carolina 29201 American Nuclear Insurers c/o Dottie Sherman, ANI Library The Exchange, Suite 245 270 Farmington Avenue Farmington, CT C6032 M & M Nuclear Consultants 1221 Avenue of the Americas New York, New York 10020 INPO Records Center Suite 1500 1100 Circle 75 Parkway
' Atlanta, Georgia 30339 Mr. W. T. Orders NRC Resident Inspector Catawba Nuclear Station Mr. R. E. Martin Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 14H25 Washington, D.C. 20555
h RCCA Positioning For Catawba 1 Cycle 7, the fully withdrawn RCCA position will vary from 222 steps withdrawn (SWD) to 230 SWD. The top of the active fuel corresponds to approximately 225 to 226 SWD.-The neutronic behavior of th cere will be affected by maintaining the RCCA's panially in the active fuel region.
The impact of the RCCAs being inserted to 222 SWD was evaluated in a 10 CFR 50.59 evaluation "RCCA Fully Withdrawn Operation at 222 steps or Above" dated October 8, 1991.
This evaluation relied c1 a Reload Safety Evaluation (RSE) perfonned by Wesunghouse on Catawba 2 Cycle 5. The conclusions of this RSE were that the various core pnysics parameters were determined to remain bounded by previous analyses.
Duke Power Company has also evaluated the impact of partially insened control rods. The following parameters were examined:
Axial Offset, Reactivity 1 ( Boron Concentration),
Peaking Fa." ors F and Fm),
o Shutdown Margin, Trip Reactivity, and Rod Worths.
The results of this analysis showed a negligible impact from positioning the RCCA's slightly in the active fuel rtgion. In' addition, Catawba 1 Cycle 6 recently repositioned their fully withdrawn RCCA position from 222 to 226 SWD. Predictions by Duke Power over-predicted the impact of withdrawing the RCCA's out of the active fuel. This implies that the impact of having the RCCA's positioned in the active fuel was being over-predicted. Therefore, since having the RCCA's partially inserted was determined to have a small effect on the important physics parameters, and the impact of the slightly inserted RCCA's was over-predicted, it can be concluded that positioning the RCCA's to 222 SWD will have a negligible effect on the neutronic behavior of the core. Additionally, it should be noted that all calculations used for predictions are performed at an intennediate rod position between 222 and 230 SWD. All safety n: lated calculations are performed at the most conservative fully withdrawn position.
Rod Ejection Pin Failure The radiological consequences of a postulated rod ejection accident must limit the offsite dose to an acceptable level. One of the assumptions for calculating the offsite dose is the number of failed pins due to depanure from nucleate boiling (DNB). The methodology documented in Duke Power Company's topical repon DPC-NE-3001 assumes that 50% of the fuel pins fail. As part of the analysis for each reload core, the number of pins experiencing DNB due to the rod ejection accident is calculated. The actual number of failed pins is then verified to be less than the 50% limit assumed in the dose analysis.
In the cycle specific analysis, a 3-D nodal code (EPRI-NODE-P) is used to calculate a 3 D assembly average power distribution. A separate code (PDQ-XD)is used to calculate the rak pin for each assembly. PDQ-XD calculates the peak pin to assembly average factor for the peak pin in each assembly. The assembly average power from the nodal code is multiplied by the pin to assembly factor for each assembly to obtain the peak pin for that assembly. To determine if a pin has exceeded the DNB limit, a thermal hydraulic a talysis is performed using the computer code VIPRE-01 to calculate a Maximum Allowable Radial Peak (MARP). If the peak pin for an assembly exceeds the MARP, it is assumed to be failed. Since the pin to assembly factor is only for the peak pin in an assembly, it is conservatively assumed that if the peak pin fails, then all 264 pins in that assembly fail.
Hence the actual number of failed pins is over predicted using this method. If the 50%
failed fuel limit is not met using this method, then a more detailed pin count can be performed using the actual pin power distributions for each assembly. The percentage of failed pins can be reduced by approximateij 20% at beginning of cycle and slightly less than 10% at end of cycle using the more detailed pin count.
Other conservatisms also exist which lead to the over predictir.g of the percentage of fuel pin failures. Two of the more significant conservatisms are as follows:
d 1)
The gas gap between the fuel pellet and the fuel cladding is assumed to close for all fuel in 0.5 seconds, and j
2)
A highly skewed top peaked power dissibution is assumed at the beginning of the rod ejection accident, and this power distribution is held constant for the duration of the accident.
Since an acceptable dose analysis is satisfied using a conservative failed pin calculation, this criteria of the ejected rod accident is clearly satisfied.
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."._ i U. SiNuclear Regulatory Commission-LJune 8,1992 -
p-t Page 4 --
bxc: R. C. Futrell-Ri L. Gill, Jr.
- M. H. Hazeltine.
a G.: B. Swindlehurst T. M. George-G. P. Horne -
K. P! Waldrop -
~ NCMPA-1
-NCEMC PMPA SREC i
- Group File: CN-801.01 Master File _(801.01) b I
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