ML13324A452
| ML13324A452 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/17/1984 |
| From: | Medford M SOUTHERN CALIFORNIA EDISON CO. |
| To: | Paulson W Office of Nuclear Reactor Regulation |
| References | |
| TAC-56067, NUDOCS 8410220039 | |
| Download: ML13324A452 (11) | |
Text
Southern California Edison Company P. 0. BOX BOO 2244 WALNUT GROVE AVENUE ROSEMEAD, CALIFORNIA 91770 M.O. MEDFORD TELEPHONE MANAGER, NUCLEAR LICENSING October 17, 1984 (213) 572-1749 Director, Office of Nuclear Reactor Regulation Attention: W. A. Paulson, Acting Chief Operating Reactors Branch No. 5 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:
Subject:
Docket No. 50-206 Increased Power Peakings San Onofre Nuclear Generating Station Unit 1 Provided below is a proposed change to the basis for Technical Specification 3.5.2 "Control Group Insertion Limits."
This change is proposed in connection with the safety evaluation for the Cycle 8 restart of San Onofre Unit 1. During the reload design and evaluation for each cycle, all appropriate safety parameters are evaluated.
Due to the extended Cycle 8 shutdown, these safety parameters have been reevaluated. This reevaluation demonstrated that core design and safety limits for the remainder of Cycle 8 will be satisfied with the present technical specifications. However, the proposed change to the basis of Specification 3.5.2 should be made for consistency with the safety evaluation. The change submitted herein is deemed not to require submittal of a License Amendment based on the provisions of 10 CFR 50.36(a), which requires inclusion of bases for technical specifications with license applications, but indicates such bases shall not become part of the technical specifications.
Therefore, your concurrence is requested to incorporate the basis change described herein.
As a result of the reevaluation discussed above, it has been determined that the FgH design maximum value (1.55) in the basis of Technical Specification 3.5.2 may be exceeded and should be increased to 1.57 to accommodate calculated increased power peakings. In addition, this higher value of FH should be increased as a function of power levels less than the 1347 MWt rated core power, to be consistent with current safety analyses of the plant.
The basis for Technical Specification 3.5.2 currently reads, in part:
"1. The initial design maximum value of specific power is-15 kW/ft.
The values of FWH and FO total associated with this specific power are 1.75 and 3.23, respectively.
6410220039 841017 A
PDR ADOCK 05000206 P
Mr. W.
October 17, 1984 A more restrictive limit on the design maximum value of specific power, FRH and FQ is applied to operation in accordance with the current safety analysis including fuel densification and ECCS performance. The values of the specific power, F H and Fq are 13.7 kW/ft., 1.55 and 2.89, respectively. The control group insertion limits in conjunction with Specification B prevent exceeding these values even assuming the most adverse Xe distribution."
The basis for Technical Specification 3.5.2. would be revised to read, in part:
"1. The initial design maximum value of specific power is 15 kW/ft. The values of FWH and F total associated with this specific power are 1.75 and 3.23, respectively.
A more restrictive limit on the design maximum value of specific power, FWH and FQ is applied to operation in accordance with the current safety analysis including fuel densification and ECCS performance. At 1347 MWt rated core power, the maximum values of specific power, FRH and FQ are 13.7 Kw/ft., 1.57 and 2.89, respectively. At partial power the FWH maximum values (limits) increase according to the following equation, FRH(P) = 1.57 [1 + 0.2(1-P)],
where P is the fraction of rated power. The control group insertion limits in conjunction with Specification B prevent exceeding these values, even assuming the most adverse Xe distribution."
The balance of the basis for Technical Specification 3.5.2 would remain as now presented in Appendix A to Provisional Operating License No.
A safety evaluation report for the F2H increase is enclosed as Attachment A. As concluded in this evaluation, there is sufficient DNBR margin to accommodate the increased FWH limit. The remaining design bases, technical specification limits, and the current accident analyses remain valid.
If you have any questions or desire additional information, please call me.
Very truly yours, cc: E. McKenna, NRC Project Manager J. 0. Ward, Chief, Radiological Health Branch, State Department of Health Services
ATTACHMENT A SAFETY EVALUATION FOR A FN INCREASE AH for Southern California Edison Company San Onofre Unit 1 Docket No. 50-206 September 1984 Editor:
J. Skaritka W-NFD 1618L:6/092084
TABLE OF CONTENTS TITLE PAGE
1.0 INTRODUCTION
AND
SUMMARY
1 2.0 NUCLEAR DESIGN EVALUATION 2
3.0 THERMAL-HYDRAULIC DESIGN EVALUATION 3
3.1 Introduction 3
3.2 Justification for Increasing FNH 4
3.3 Conclusion 4
4.0 ACCIDENT EVALUATION 5
4.1 Introduction 5
4.2 Non-LOCA Evaluation 5
4.3 LOCA Evaluation 6
5.0 REFERENCES
7 1618L:6/092084
1.0 INTRODUCTION
AND
SUMMARY
This report presents a safety evaluation for San Onofre Unit 1 operation N
with an increased FAH limit as a Basis for the Technical Specification 3.5.2, control Rod Insertion Limits.
The increase in the 100% rated power F Nlimit from 1.55 to AH 1.57 is needed to accommodate calculated increased power peakings due to a 32 month Cycle 8 shutdown for plant modifications and two separated RCCA rodlets stuck in fuel assembly guide thimbles.
The incidents analyzed and reported in the FSA(1) which could potentially be affected by changes described in this report have been evaluated and are discussed in later sections.
The result of the evaluation/analysis described herein lead to the following conclusions:
- 1. The proposed changes to the Section 3.5.2 Basis do not impact the other design/safety bases used for the Cycle 8 RSEs(2)(3) which were submitted for NRC review(4)
- 2. There is sufficient DNBR margin to accommodate the reduction in margin resulting from the increased FNH limit. The remaining design bases, technical specification limits and the current non-LOCA and LOCA analyses remain valid.
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2.0 NUCLEAR DESIGN EVALUATION The proposed technical specification basis changes do not impact the other nuclear design bases used to evaluate the Cycle 8 RSEs(2 )(3 )
The standard calculational methods described in the "Westinghouse Reload Safety Evaluation Methodology"(5) continue to apply. As is current practice, each reload core design is evaluated to assure that design and safety limits are satisfied according to this reload methodology.
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3.0 THERMAL-HYDRAULIC DESIGN EVALUATION 3.1 Introduction The proposed San Onofre Unit I technical specification basis changes which impact DNBR evaluation is the value of FN determined from the following equation:
FN L
F
< F (1 + 0.2 (1-P))
N where F
AH = measured radial peaking factor with appropriate uncertainties.
FL = peaking factor limit at 100% Rated Power AH P = fractional core power level at less than 100% Rated Power or,
= 1.0 at greater than or equal to 100% Rated Power.
L The radial peaking factor limit (FAH) increase from 1.55 to 1.57 has a direct impact on DNBR calculations.
The core limits of the Technical Specification Figure 2.1.1 include a restriction that the average enthalpy at the vessel exit must be less than the enthalpy of saturated liquid to assure the proportionality between vessel AT and core power. The exit enthalpy restriction is more limiting than DNBR at low heat fluxes and is independent of radial peaking factor as shown in the following relation.
h
=h. + 9 < h out in G sat where hout average coolant enthalpy at vessel exit (BTU/lb M) 1618L:6/092084 3
10 h.i = vessel inlet coolant enthalpy (BTU/lb M)
Q
= total core power (BTU/hr)
G
= total core coolant flow (lb m/hr) 3.2 Justification for Increasing NF In Cycle 8, 4.4% in DNBR margin is available due to pitch reduction and adverse design axial power shape.(3) This available DNBR margin is sufficient to accommodate the reduction in DNBR margin resulting from increasing the 100% rated power FAH from 1.55 to 1.57.
The core limit curves (Technical Specification Figure 2.1.1) remain unchanged.
3.3 Conclusion In summary, the effect of increasing the 100% rated power F N from 1.55 to 1.57 is offset by a reduction in DNBR AH margin such that the existing core DNB limits (Overtemperature AT) and non-Overtemperature AT trip accident analyses remain valid. The FSA DNBR design basis is met with the increased N
FAH 1618L:6/092084 4
4.0 ACCIDENT EVALUATION 4.1 Introduction This section summarizes the effects of the increased FN AH limit for San Onofre Unit 1 on the FSA Chapter 14 non-LOCA and LOCA analyses.
4.2 Non-LOCA Evaluation The methods used for accident evaluation are described in Reference 5 and are the same as those applied to previous San Onofre reload safety evaluations.
To ensure adequate core protection, the Reactor Core Thermal and Hydraulic Safety Limits were reevaluated due to the increased FN FH limit. Since this evaluation determined that the reactor core safety limits are not changed, the variable low pressure setpoint equation a given in the technical specification will not change. Therefore, no reanalysis is required for transients requiring the variable low pressure trip for protection.
The overall system transient response is not affected by the increased F N limit. Rather, the effect of the increased AH FAH is accounted for in the DNBR calculations.
is As noted in Section 3.0, the penalty incurred as a result of the increased F N is offset by a reduction in available DNBR margin. Therefore, the current non-LOCA accident analyses remain valid and no reanalysis is required.
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4.3 LOCA Evaluation The current large break LOCA FSAR analysis continues to establish conformity with the requirements of (Reference 6) for an increase in N
enthalpy rise peaking factor (F H) from 1.55 to 1.57. The analysis was performed assuming a core average heat flux of 4.764 kw/ft based on a core licensed power of 1347 MW produced by 180 rods in each of 157 assemblies. The hot rod power distribution is modeled in LOCTA-R2 (Reference 7) as the limiting chopped cosine shape having a maximum heat flux of 13.74 kw/ft. The total hot rod peaking factor (FQ T), the ratio of the hot rod maximum heat flux to the core average heat flux, is therefore 2.89.
Keeping the core average coolant enthalpy rise constant while increasing F N represents an increase in the hot channel coolant enthalpy AH rise.
This results in a change to the hot rod chopped cosine power distribution used in calculating the large break LOCA peak clad temperature. The change is an increase in the average heat flux of the hot rod.
The current large break LOCA analysis of record has assumed an enthalpy rise peaking factor of 1.75. Therefore, for an F T of 2.89, the analysis remains applicable and conservatively bounds any enthalpy rise peaking factor less than 1.75.
The hot rod axial power distribution assumed in the small break LOCA analysis of record continues to be applicable for an F T of 2.89 and an enthalpy rise peaking factor of 1.57.
In summary, the current large and small break LOCA analyses of record remain applicable for an enthalpy rise peaking factor of 1.57, and no further LOCA analyses or evaluations are required to support this technical specification change.
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5.0 REFERENCES
- 1. Docket Number 50-206, "San Onofre Nuclear Generating Station,Unit 1, Part 2, Final Safety Analysis".
- 2. Skaritka, J., Editor, "Reload Safety Evaluation - San Onofre Unit 1, Cycle 8 -
Revision 1," October 1980.
- 3. Skaritka, J., Editor, "Reload Safety Evaluation -
San Onofre Unit 1 Cycle 8 -
Revision 2", April, 1981.
- 4. Letter from K. P. Baskin (SCE) to D. N. Crutchfield (NRC);
Subject:
Amendment 95 to San Onofre Unit 1 Operating License; Docket No..50-206; April 15, 1981.
- 5. Bordelon, F. M., et. al., "Westinghouse Reload Safety Evaluation Methodology", WCAP-9272 (Proprietary), March 1978.
- 6.Section IV.A and B, "Criteria for Emergency Core Cooling Systems for Light Water Power Reactors", Federal Register, Vol. 36, No. 125, June 29, 1971.
- 7. Bezella, W. A., Caso, C. L., and Spencer, A. C., "LOCTA-R2 Program:
Loss of Coolant Transient Analysis", WCAP-7437-L, January, 1970.
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