ML19338C781
| ML19338C781 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/11/1980 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19338C783 | List: |
| References | |
| NUDOCS 8009050240 | |
| Download: ML19338C781 (5) | |
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SACRAMEflTO MUtilCIPAL UTILITY DISTRICT O 6201 S Street, Box 15830. Sacramento, Cal;fornia 95313, (916) 452-3211 August 11, 1980 i
Director of Nuclear Reactor Regulation Attention:
Mr. Robert W. Reid, Chief Operating Reactors, Branch No. 4 j
U. S. Nuclear Regulatory Commission Washington, D. C.
20555 i
Docket No. 50-312 Proposed Amendment No. 70 Rancho Seco Nuclear Generating Station, Unit No. I
Dear Mr. Reid:
In accordance wi th 10 CFR 50.59, the Sacramento Municipal Utili ty District proposes to amend its operating license, DPR-54, for Rancho Seco Nuclear Generating Station No.
1, by submitting Proposed Amendment No. 70 on August 18, 1980.
Today, we are submitting forty (40) copies of Proposed Amendment No. 70 which shows the changes we are proposing.
Under separate cover, we will be providing payment for this submittal as required per 10 CFR 170.
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We. request your expeditious review and issue of this Proposed Amend-ment to our license to assist in resolving our concern that we not inadvertently challenge the Reactor Protection System due to spurious Reactor Coolant System flow oscillations.
This revision will change the safety analysis assumed " flux to flow" ratio f rom 1.05 to 1.08.
Such a change is justified as a result of re-analysis performed by our fuel supplier which takes credit for the decrease in reactor core bypass flow due to the installation of 52 lumped burnable poison assemblies within the core.
The resulting increase in ef fective core coolant flow has been incorporated into the safety analysis and demonstrates that Rancho Seco can operate at the design pewer level and meet all transient and safety considerations successfully with the new power to flow ratio.
In support of this request, we are submitting revised text for Sections 6 and 7 in the Cycle 4 Reload Report, BAW-1560.
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Sincerely,
.WC J. J. Mattimoe i
Assistant General Manager and Chief Engineer JJM: RUC:j r Sworn to and subscribed before me tnis 7/' day of August'[/
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4 6.
TIIERMAL-llYDRAULIC DESIGN The incoming batch 6 fuel is hydraulically and geometrically similar to the fuel remaining in the core from previous cycles.
The thermal-hydraulic design evaluation supporting cycle 4 operation utilized the methods and models de-scribed in references 2, 3, and 7 except for the core bypass flow and the in-clusion of retainers to provide positive holddown of burnabic poison rod assem-blies (3??As) and neutron sources.
The maximum core bypass flow due to the removal of all orifice rod assemblies (ORAs) in cycle 3 increased to 10.4%.
For cycle 4 operation, 52 BPRAs will be inserted, leaving 56 vacant fuel assemblies an1 resulting in a decrease in cal-culated maximum core bypass flow to 8.3%.
The BPRA retainers introduce a small DNBR penalty, as discussed in reference 1.
Reactor core safety limits have been re-evaluated based on the insertion of these BPRAs with retainers and increased core flow.
The cycle 3 and 4 maximum design conditions and sig-nificant parameters are shown in Table 6-1.
The increase in core flow more than compensates for the decrease in DNBR due to the BPRA retainers so that the cycle 3 analysis is conservative and applicable to cycle 4.
A flux / flow trip setpoint of 1.08 has been established for cycle 4 operation.
This setpoint value maintains a DNBR margin which is greater than 10%, relative to the design minimum DNBR of 1.30 (BAW-2).
The previous setpoint value of 1.05 had initially leen established for first core operation and has been verified as conservative and, taerefore, applicable for succeeding cycles.
In response to reference 8, B&W has com.Mcted to prepare a topical report ad-dressing the potential for and effects of fuel rod bow.
In addition, B&W has I
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submitted an interim rod bow penalty evaluation procedure for use until the topical report is completed and reviewed.
The rod bow penalty applicabic to cycle 4 was calculated using the interim rod bow penalty evaluation procedure. As in the previous cycle the burnup is based on the maximum fuel assembly burnup of the batch that contains the fuel assem-bly with the maximum radial x local peak.
For cycle 4 this burnup is 14,537 mwd /mtU, a batch 6 fuel assembly.
The calculated penalty using this procedure is less than 0.8%.
Utilizing the 1%'DNB credit for the flow area reduction factor, the actual penalty applied to the DNB calculate ns is zero.
7.
ACCIDENT AND TRANSIENT ANALYSIS 7.1.
Cencral Safety Analysis Each FSAR accident analysis has been examined with respect to changes in cycle 4 parameters to determine the effect of the cycle 4 reload and to cusure that thermal performance during hypothetical transients is not degradcd.
The effects of fuel densification on the FSAR accident results have beer, evaluated and are reported in BAW-1393.
Since the cycle 4 parameters are conservative with re-spect to the reference 7 report, the conclusions in that reference are still valid.
Improved fuel utilization and the inherent increase in core average burnup ex-perienced in cycle 4 have resulted in a higher plutonium-to-uranium fission ratio than that used in the FASR. A stwi of the major FSAR Chapter 14 acci-dents using the cycle 4 iodine and nob].. gas inventories concluded that the thyroid and whole body doses were well below the 10 CFR 100 limits.
7.2.
Accident Evaluation The key parameters that have the greatest effect on determining the outcome of a transient can typically be classified in three major areas: core thermal, thermal-hydraulic, and kinetics parameters, including the reactivity feedback coefficients and control rod worths.
Core thermal properties used in the FSAR accident analysis were design operat-ing values based on calculational values plus uncertainties.
First-core values of core thermal parameters and subsequent fuel batches are compared to those used in cycle 4 analyses in Table 4-t The cycle 4 thermal-hydraulic maximum design conditions are compared to cycle 3 values to Table 6-1.
These parameters are common to all the accidents considered in this report.
A com-parison of the key kinetics parameters from the FSAR and cycle 4 is provided
in Table 7-1.
Cycle 4 parameters include the effects of removing the orifice rod assemblies.
Additionally, all accident analyses, and their related evalu-ations, continue to be valid with respect to the new flux / flow setpoint of 1.08.
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