ML20044C323
| ML20044C323 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 03/10/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044C321 | List: |
| References | |
| NUDOCS 9303220181 | |
| Download: ML20044C323 (6) | |
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4 UNITED STATES j
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 4001 o
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION e
RELATED TO AMENDMENT NO. 57 TO FACILITY OPERATING LICENSE NPF-68 f
AND AMENDMENT NO. 36 TO FACILITY OPERATING LICENSE NPF-81 GEORGIA POWER COMPANY. ET AL.
V0GTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 DOCKET NOS. 50-424 AND 50-425
1.0 INTRODUCTION
By [[letter::ELV-03832, Application for Amends to Licenses NPF-68 & NPF-81,revising TS SR 4.5.2.b by Changing Frequency for Verifying That ECCS Piping Is Full of Water from Once Per 31 Days to Once Per 6 Months|letter dated September 17, 1992]], Georgia Power Company, et al. (the licensee), proposed license amendments to change the Technical Specifications (IS) for the vogtle Electric Generating Plant (Vogtle or the facility), Units 1 and 2.
The proposed amendments would revise the time constant (7 ) used in 3
the lag compensator for delta temperature (AT) in the overtemperature delta temperature (OTAT) and overpower delta temperature (OPAT) reactor trip i
function setpoint equations in footnotes 1 and 3 of TS Table 2.2-1 from 0 seconds to 2.0 seconds. The K term of the OPAT equation in footnote 3 would i
4 also be changed from 1.08 to 1.095.
By letter dated February 12 and 25, 1993, the licensee provided additional information in support of the application for amendments.
The additional information does not affect the NRC's proposed finding of no significant hazards considerations as published in the Federal Reaister (57 FR 47135)
October 14, 1992.
2.0 BACKCROUND The OTAT reactor trip provides core protection to prevent departure from nucleate boiling (DNB) for all combinations of pressure, power, coolant -
temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), and pressure is within the range between the pressurizer-high and low pressure reactor trip setpoints.
The OPAT reactor trip protects fuel integrity (e.g., prevents fuel pellet melting and excessive -- less than 1% -- cladding strain) under overpower conditions, limits the required range for the OTAT reactor trip, and backs up the high neutron flux reactor trip. The OPAT reactor trip also provides protection to mitigate the consequences of steam line break (SLB) accidents.
In addition to their reactor trip functions, OTAT and OPAT provide a signal to generate a turbine runback before the reactor protection system reaches a reactor trip.setpoint.. The turbine runback reduces turbine power and reactor 9303220181 930310 gDR ADOCK 05000424 l
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-i power to alleviate the OTAT or OPAT condition and preclude' the need for a reactor trip.
Two events prompted the licensee to request changes to the OTAT and'0 PAT reactor trip functions First,' a flow-phenomenon in the upper plenum of the reactor. vessel' has -
been recently observed. This phenomenon is manifested as a sudden rise-in temperature in one of the reactor coolant system (RCS) hot legs with a concurrent decrease in temperature in the adjacent hot leg..The condition occurs randomly, lasts for a'short duration, and temperatures:
then return to their previous values.
In some observations;at:Vogtle, the temperature rise has been sufficient:to-cause.a turbine runback signal from the affected hot leg. This runback signal is jonsidered to be spurious and not indicative of an actual overtemperature.or overpower condition.
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Second, and by separate action, the licensee-has proposed amendments to' increase core power from 3411 megawatts = thermal- (Mwt) to' 3565 Nwt.
Extrapolations' based upon the magnitude of the flow phenomenon' at Vogtle's current-power levels indicate that additional lmiargin between the setpoint and the operating conditions for the upratdd power may be needed to avoid spurious turbine runback signals. To gain this' margin.
to avoid spurious runback signals, the licensee proposes the.following TS changes:
Change 7 from 0 to 2 seconds in both~the OTAT and OPAT reactor' 3
trip functions. 7 is the time conttant in the-lag compensator-3 for measured AT. This proposed change-would dampen the response L of the system.to any temperature spike-of short duration.
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Change K from 1.08 to 1.095 in the 0 PAT reactor trip function. K l
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defines a maximum power excursion limit. This would provide addition operating margin to compensate for-the effect of the-upper head flow phenomenon at the.uprated power conditions.
i 3.0 EVALUATION The impact of the changes to the OTAT and OPAT reactor trip functions has been assessed by the effect of the changed OTAT and OPAT setpoints on the events in 1
the Vogtle Final Safety Analysis Report (FSAR) that rely upon these trips for protection. The bases used in determining the proposed OTAT and OPAT reactor-trip setpoints were also reviewed.
3.1 Events That Rely Upon OTAT Reactor Trip Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal at -
Power (FSAR Section 15.4.21 Uncontrolled RCCA bank withdrawal-at-power results in an increase in -
core heat. flux. Unless terminated by manual or automatic action, the 7
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. power mismatch and resultant coolant temperature rise could eventually result in DNB. To prevent damage to the fuel clad, the reactor protection system is designed to terminate any such transient before the departure from nucleate boiling ratio (DNBR) falls below its safe limit.
Currently, the high neutron flux and OTAT reactor trips. provides protection over all reactivity insertion rates.
To address the effect of the increase in 7 from 0 to 2 seconds, the 3
licensee reevaluated the relevant events currently presented in the FSAR. The revised analyses were performed using the same methods used in the FSAR analyses. The results show that events protected by a.
reactor trip from the high neutron flux signal would be unaffected by the increase in 7, and that the events protected by the OTAT reactor 3
trip would be only slightly affected.
For all events, the minimum DNBR would remain above the safety analysis limit. Thus, the conclusions presented in the FSAR remain valid.
Inadvertent Openina of a Pressurizer Safety or Relief Valve (FSAR Section 15.6.1)
Events that result in a decrease in reactor coolant inventory include an inadvertent opening of a pressurizer safety or relief valve.
Of these two events, the inadvertent opening of a pressurizer safety valve causes the more severe core conditions from an accidental depressurization of the RCS. The pressurizer safety valve opening is worse because it relieves about twice the steam flow of a relief valve and, thus, results.
in a much faster depressurization upon opening.
Currently, the OTAT and the pressurizer low pressure reactor trips provide adequate protection, and maintain DNBR above the safety limit.
The licensee's reevaluation addressed the effect of the increase in 7 3 from 0 to 2 seconds and the RCS depressurization event.
The analyses were performed using the same methods used in the FSAR analyses.
The res'ults demonstrate that the DNBR limit would continue to be met.
Therefore, the conclusions presented in the FSAR remain valid.
Chemical and Volume Control System (CVCS) Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant (FSAR Section 15,4.6)
Reactivity can be added to the core by feeding primary grade water into the RCS via the CVCS.
Boron dilution is a manual operation under strict administrative controls with procedures calling for a limit on the rate and duration of dilution. The CVCS is designed to limit the potential rate of dilution.
If a CVCS malfunction results in a dilution of the RCS boron concentration, the operator is alerted by instrumentation and alarms in time to correct the situation before a significant loss of shutdoe ~ rgin.
During full power operation with the reactor in manual control, the operator would be alerted to an uncontrolled boron dilution of the reactor coolant by an OTAT reactor trip.
(This is the only situation
3 associated with the postulated accident that relies on the OTAT reactor trip). Without the proposed OTAT change, the current FSAR analysis shows that at least 16.9 minutes would be available for operator action after the reactor trip before the loss of' shutdown margin.. In the licensee's reevaluation for the OTAT changes, with 7 increased from 0 3
to 2 seconds, the change in time of reactor trip is less than 3 seconds.
A change of such small magnitude is insignificant relative to' the total time available for operator action following the trip before.a loss of-shutdown margin. Moreover, the criterion for operator action time'(15 4
minutes) would continue to be met and the conclusions in the FSAR would remain valid.
Steam Generator Tube Rupture (FSAR 15.6.3)
The steam generator tube rupture (SGTR) accident is assumed to result from the complete severance of a single steam generator tube. The accident leads to contamination of the main steam system due to leakage of radioactive coolant from the RCS.
In the event of a coincident loss of offsite power, the steam dump valves would automatically close to protect the condenser.
Steam generator pressure would increase. rapidly, resulting in the discharge of steam to.the atmosphere through the steam generator power-operated relief' valves (and the safety valves if their -
setpoints are reached).
The mass releases are used for evaluating offsite radiation exposures.
The licensee's evaluation addressed the increase in 7 from 0 to'2 3
seconds.
The delay in the time of the reactor trip had no effect on the SGTR response since the analysis conservatively assumes a 0-second delay between reaching the OTAT setpoint and the reactor trip.
For the SGTR event, a delay in the reactor trip signal would not increue the amount of steam released through the atmospheric relief valves and, therefore, would not increase the offsite radiation exposure. Therefore, the conclusions in the FSAR remain valid.
3.2 Events That Relv Upon OPAT Reactor Trio Steam Line Break Coincident With a Control Rod Withdrawal (FSAR Section 15.4.9)
A SLB in the vicinity of the turbine impulse pressure transmitters or the excore detectors may expose equipment used in the rod control to an adverse environment.
If the associated cabling and connections are not properly qualified, then the potential would exist for steam to impinge:
upon this equipment and cause a control system malfunction. The resuling malfunction could initiate a control rod withdrawal during the-SLB accident.
Currently, the OPAT reactor trip provides protection from this postulated accident and prevents DNBR from decreasing below the safety limit.
The licensee reevaluated this accident using the increase in 7 from O' 3
to 2 seconds and the change in K from 1.08 to 1.095. The licensee used acceptable methods for the analysis. The results demonstrated that the
. DNBR limit would continue to be met. Therefore, the FSAR conclusions remain valid.
Analysis of Steam Line Break With Super-Heat (WCAP-ll285)
The licensee has previously analyzed four cases of a SLB with super-heated steam for which a reactor trip occurs from an OPAT signal.
In each of these cases, the total length of the transient is about 1800 seconds and the trip occurs early in the transient (before 30 seconds).
The licensee used acceptable methods for these analyses.
Based on the analysis performed for the SLB coincident with a control rod withdrawal (see above), the proposed modification to the OPAT reactor function would not delay the reactor trip by more than a few seconds in the analyses of SLB with super-heat. A delay of such short magnitude would not result in any significW change in the overall profile of the mass and energy releases cr super-heated conditions due to the extended length of the transient. Therefore, the analyses for SLB super-heat remain valid.
3.3 ktpoint Bases The methods used by Westinghouse to calculate the OTAT and OPAT setpoints are provided in topical report WCAP-8745-P-A, " Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," dated March 1977. This topical report was approved by the NRC April 17, 1986. The NRC staff verified that the proposed changes to 7 and K for Vogtle were Completed in aCCordance with the methods 3
4 described in WCAP-8745-P-A.
WCAP-8745-P-A also addresses the uncertainties in the OTAT and OPAT reactor trip functions, including the calibration and instrumentation channel error.
The NRC staff accepted these uncertainties and error values during its approval of WCAP-8745-P-A.
Based on the above review, the staff finds that the proposed changes to the OTAT and OPAT reactor trip functions are based upon analyses performed in accordance with NRC approved methods. The licensee's evaluations were performed with the same models previously used in the FSAR safety analyses.
Results of the evaluation show that existing safety criteria are met.
Specifically, the analyses demonstrate that the limits for DNBR and fuel temperature would not be exceeded for the revised OTAT and OPAT reactor trip functions. Therefore, the NRC staff finds that the proposed changes are acceptable.
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4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Georgia State official was notified of the proposed issuance of the amendments. The State official-had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements 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 increas.e 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 previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no r
public comment on such finding (57 FR 47135, dated October 14,.1992).
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 he 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.
Principal Contributors:
D. Hood R. Skokowski H. Balukjian P. Balmain Date: March 10, 1993 i
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