ML20217H940
| ML20217H940 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 03/27/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217H922 | List: |
| References | |
| NUDOCS 9804060125 | |
| Download: ML20217H940 (3) | |
Text
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I UNITED STATES s
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. Samaa ana1
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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGUI.ATION
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RELATED TO AMENDMENT NOS.57 AND 43 TO l
l FACILITY OPERATING LICENSE NOS. NPF-87 AND NPF-89 l
TEXAS UTILITIES ELECTRIC COMPANY COMANCHE PEAK STEAM ELECTRIC STATION. UNITS 1 AND 2 DOCKET NOS. 50-445 AND 50-446 l
1.0 INTRODUCTION
. By application dated October 24,1997 (TXX-97228), Texas Utilities Electric Company (TU l
Electric /the licensee) requested changes to the Technical Specifications (Appendix A to Facility Operating License Nos. NPF-87 and NPF-89) for the Comanche Peak Steam Electric Station, l
Units 1 and 2. The proposed changes we.ild revise core safety limit curves and Overtemperature N-16 reactor trip setpoints based on analyses of the core configuration for CPSES Unit 1, Cycle 7. The specific changes are to TS Figure 2.1-1a, Unit 1 Reactor Core Safety Limits and TS Table 2.2-1, " Reactor Trip System Instrumentation Trip Setpoints," Units 1 and 2. These changes are required due to the change in core configuration for CPSES Unit 1, Cycle 7 refueling.
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2.0 BACKGROUND
The core safety limits are the set of points of Thermal Power, Reactor Coolant System (RCS) pressure, and average temperature below which (1) the calculated Depadure from Nucleate i
Boiling Ratio (DNBR) is no less than the safety analysis limit value, or (2) the average enthalpy at l
the vessel exit is less than the enthalpy of saturated liquid. The significant parameters used to determine these lines include the RCS flow rate, the design nuclear enthalpy rise hot channel factor (FW, the cycle-specific reference axial power shape, and the cycle-specific core l
configuration.
To avoid exceeding the core safety limits during operation, the Overtemperature N-16 reactor trip initiates a reactor trip before the limits are exceeded, providing DNB protection from events that impact power, pressure, temperature, or axial power shape. The setpoint for the N-16 trip is continuously calculated for each loop by analog circuitry by solving the setpoint equation that includes the trip reset function term, f(AI).
The actual core power is measured on each loop with an N-16 power meter and used as a comparison to the calculated trip setpoint. Should an axial pcwer distribution occur during operation that is more severe than the reference axial power shape, the trip reset function term of the setpoint equation, reduces the trip setpoint to account for the resultant effects. The range over which the Overtemperature N-16 trip setpoint is calculated is bounded by the pressurizer pressure low, the pressurizer pressure high and the overpower reactor trip setpoints. The power / temperature range that the Overtemperature N-16 trip must provide DNB protection is further limited by the operation of the Main Steam Safety Valves.
9804060125 980327 DR ADOCK 050 4j5
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~ 3.0 EVALUATION l
The Unit 1 Cycle 7 core configuration consists of 192 fuel assemblies manufactured by Siemens
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. Power Company and 1 fuel assembly manufactured by Westinghouse Electric Company. The mixed-core penalty for Cycle 7 is the same as Cycle 6. The licensee also calculated the Unit 1,-
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Cycle 7 axial power distributions to be more skewed toward the top half of the core than those calculated for Unit 1, Cycle 6. Therefore, it was necessary for the licensee to change the f(AI).
j trip function of the Overtemperature N-16 trip setpoint to reflect the Unit 1, Cycle 7 core configuration.
The licensee indicated that they used NRC-approved methodologies as stated in TS 6.9.1.6.b '
Items 9,10,11,12,13, and 14 to determine the new reactor core safety limit curves for Unit 1, TS figure 2.1-1a. The resultant core safety limit curves were evaluated by NRC-approved j
l-analytical methods to determine the appropriate values of K,, K, and K, coefficients of the i
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. overtemperature setpoint and the f(AI) trip reset function as noted below:-
l in TS Table 2.2-1, Note 1 for the Overtemperature N-16 Trip Setpoint, the following terms have l-been changed for Unit 1:
K revised from 0.0173 to 0.0148/'F 2
K revised from 0.000890 to 0.00080/psig.
l-3 qi-q. range from -65% and +4.6% to -65% and +5.0%
l Automatic Overiemperature N-16 setpoint reduction for each percent that the magnitude i
of qi-q, exceeds +5.0% (current value +4.6%) is decreased from 3.04% to 2.15%.-
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In TS Table 2.2-1, Note 2 for the Overtemperature N-16 Allowable Value, the maximum amount by which the Trip Setpoint is allowed to exceed the computed Trip Setpoint, is decreased from 3.64% to 1.72%.
The licensee determined that the CPSES Final Safety Analysis Report (FSAR), Chapter 15 event i
l most affected by the change in the Overtemperature N-16 trip setpoint is Section 15.4.2, " Rod l~
ithdrawal at Power." The licensee reanalyzed the event using NRC-approved methodologies W
l listed in TS 6.9.1.6.b Items 9,10,11,12,13, and 14 and determined that all relevant acceptance criteria were satisfied.
The licensee has requested to change TS Figure 2.1-1a, Unit 1 Reactor Core Safety Limits and TS Table 2.2-1, " Reactor Trip System Instrumentation Trip Setpoints," Units 1 and 2 to accommodate changes due to the designed core configuration for Unit 1, Cycle 7. The staff has reviewed the changes to verify that all safety criteria have been satisfied. The proposed revisions to the safety limit curves and the trip setpoints, and the use of NRC-approved methodologies will provide assurance of DNB margin. Therefore, the staff finds the licensee's requested changes acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Texas State official was notified of the l
proposed issuance of the amendments. The State official had no comments.
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5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previcusly issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (62 FR 61847). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
PrincipalContributor: M. Gamberoni Date:
March 27, 1998 i
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