ML20151W307
| ML20151W307 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 09/08/1998 |
| From: | CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20151W272 | List: |
| References | |
| NUDOCS 9809150255 | |
| Download: ML20151W307 (2) | |
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.LAR No. 98-0006
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'Page 8 of 9 2.1 SAFETY LIMITS BASES 2.1.1 AND 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the
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cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime would result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and l
Reactor Coolant Temperature and Pressure have been related to DNB using critical heat flux (CHF) correlations. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The B&W-2 and BWC CHF correlations have been developed to predict DNB for axially uniform and non-uniforrn heat flux distributions. The B&W-2 correlation applies to Mark-B fuel and the BWC correlation applies to all B&W fuel with zircaloy gLM5 spacer grids. The minimum value of the DNBR during steady state operation, normal j
operational transients, and anticipated transients is limited to 1.30 (B&W-2) and 1.18 (BWC). The value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR equal to or greater than the correlation limit is predicted for the maximum possible thermal power 112% when the reactor coolant flow is 380,000 GPM, which is approximately 108% of design flow rate for four operating reactor coolant pumps. (The minimum required measured flow is 389,500 GPM). This curve is based on the design hot channel factors with potential fuel densification and fuel rod bowing effects.
The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum allowable control rod withdrawal, and form the core DNBR design basis.
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PDR ADOCK 05000346 P
PDR DAVIS-BESSE, UNIT 1 B 2-1 Amendment No. I1,33,91,123, 149,189,
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Design Features 5.0 5.0 DESIGN FEATURES 5.1 Site Location The Davis-Besse Nuclear Power Station, Unit Number 1, site is located on Lake Erie in Ottawa County, Ohio, approximately six miles northeast from Oak Harbor, Ohio and 21 miles east from Toledo, Ohio.
. The exclusion area boundary has a minimum radius of 2400 feet from the center of the plant.
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5.2 (Deleted) t 5.3 Reector Core i
5.2... ael Assemblies f
i The reactor core shall contain 177 fuel assemblies. Each assembly shall consist of a matrix of zircaloyE or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO ) as fuel material. Limited substitutions of zirconium alloy or 2
stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applich k NRC staff approved codes and methods and shown by tests or t
analyses to comply with all ioel safety design bases. A limited number oflead test assemblies that have not completed representative testing may be placed in non-limiting core
- regions, j
l 5.3.2 Control Rods The reactor core shall contain 53 safety and regulating control rod assemblies and 8 axial power shaping rod (APSR) assemblies. The nominal values of absorber material for the safety and regulating control rods shall be 80 percent silver,15 percent indium and 5 percent cadmium. The absorber material for the APSRs shall be 100 percent Inconel.
5.4 (Deleted) 5.5 (Deleted) 5.6 Fuel Storage 5.6.1 Criticality 5.6.1.1 The spent fd storage racks are designed and shall be maintained with:
I A K,n quivalent to less than or equal to 0.95 when flooded with unborated water, a.
e which includes a conservative allowance of 1% delta k/k for calculation uncertainty.
(continued)
DAVIS-BESSE, UNIT 1 5-1 Amendment No. 170,179,204,
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