ML20138Q134
| ML20138Q134 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 12/13/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20138Q120 | List: |
| References | |
| NUDOCS 8512270064 | |
| Download: ML20138Q134 (5) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 73 AND 59 TO i
FACILITY OPERATING LICENSE NOS. NPF-4 AND NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE NORTH ANNA POWER STATION, UNITS NO. 1 AND NO. 2 DOCKET NOS. 50-338 AND 50-339 I
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Introduction:==
By latter dated February 7,1985 (Serial No. 666), the Virginia Electric and Power Company (the licensee) requested a change to the Technical Specifications (TS) for the North Anna Power Station, Units No. I and No. 2 (NA-1&2).
Specifically, the proposed change would allow operation with a positive Moderator Temperature Coefficient (MTC) at reduced power levels. The proposed change would allow greater flexibility in core designs at NA-1&2 for future l
cycles.
Discussion:
The proposed change would also minimize the necessity of having the control rods significantly inserted in the core during initial startup and the potential for operating restrictions due to the delta flux limits associated with j
constantagaloffsetcontrol. The licensee proposed a TS which allows a MTC of +6 X 10 AK/K/*F for power levels below 70 percent of rated power and a i
zero coefficient for power levels 70 percent and above.
The power dependent MTC was proposed to minimize the effect on accidents initiated from high power levels.
The present NA-1&2 TS does not allow the reactor to be critical unless the MTC is negative, except during physics tests.
Design calculations for recent NA cycles indicate that a positive MTC may occur at the beginning of cycle for hot zero power conditions with all rods removed from the core.
While control rod insertion may be used to make the coefficient negative, l
this makes startup more complicated and takes longer.
As power level increases, the allowed average coolant temperature becomes higher and the MTC becomes more negative.
Also the boron concentration is reduced as xenon builds into the core.
Thus a positive MTC is not needed as full power is approached. As fuel burnup is achieved, borcn is further reduced and the MTC becomes more negative over the entire operating power range.
It is expected that the MTC would be positive only for low powers at beginning of cycle.
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2-Evaluation:
The licensee reanalyzed those NA-1&2 Updated Final Safety Analysis Report (UFSAR) Chapter 15incidentswhichweresensitivetoafginumornearzero MTC.
All the reanalysis was done with a MTC of 6 X 10 AK/K/*F.
No credit was taken for change in MTC due to increases in temperature or power.
In general, the reanalysis was based on the assumptions and methods used for the UFSAR Accidents. The accidents not reanalyzed include those resulting in excessive heat removal from the reactor coolant system (for which a large negative MTC is limiting) and those which experience heatup following a reactor trip (which are not sensitive to the MTC).
The following transients were found to be not affected by a positive moderator coefficient.
Rod Cluster Control Assembly Misalignment Startup of an Inactive Reactor Coolant Loop-Excessive Heat Removal Due to Feedwater System Malfunctions Excessive Load Increase Loss of Normal Feedwater, Loss of Offsite Power to Station Auxiliaries Accidental Depressurization of the Reactor Coolant System Rupture of a Main Steam Pipe / Accidental Depressurization of the Main Steam System Spurious Operation of Safety Injection Rupture of a Main Feedwater Pipe Loss of Coolant Accident Transients Sensitive to a Positive Moderator C' efficient o
Uncontrolled Doron Dilution Boron dilution at power causes an increase in power and reactor coolant system temperature if the reacter is in manual control. With a positive MTC, the temperature increase would result in adding additional reactivity and l
increasing the severity of the transient.
However, this incident is less severe than the rod withdrawal at power and therefore is bounded by the results of that analysis.
Control Rod Withdrawal from a Subcritical Condition l
l A control rod assembly withdrawal incident when the reactor is subcritical l
results in an uncontrolled addition of reactivity leading to a power l
excursion. The analysis showed that the peak heat flux, peak coolant l
temperature and thermal power did not exceed normal full power values.
Since the heat flux does not exceed the normal full power value and remains bounded by the UFSAR results, the conclusions presented in the UFSAR are stil! valid.
a.
Uncontrolled Control Rod Assembly Withdrawal at Power An uncontrolled control rod assembly withdrawal at power produces a mismatch in steam flow and core power, resulting in an increase in reactor coolant temperature. ApositiveMTCincreasesthepowggmismatchandreducesmargin to DNS. The event was analyzed with a +6 X 10 A/K/K'F MTC, even though a positive MTC would be allowed only below 70 percent of power. The minimum DNBR was found to be greater than the limit of 1.3 for the earlier range of reactivity insertion values.
Thus it was demonstrated that the positive MTC did not lower the DNBR associated with control rod assembly withdrawal at power to below the design limit.
Loss of Reactor Coolant Flow The most severe loss of flow transient is caused by simultaneous loss of electric power to all three reactor coolant pumps.
For the case reanalyzed, the reactor coolant average temperature increase was less than 3'F above the initial value. Analysis with the positive MTC showed that the minimum DNBR was greater than 1.30.
Thus it was demonstrated that the results of the complete loss of flow transient remain above the 1.30 limit for DNBR.
Locked Rotor i
The UFSAR shows that the most severe locked rotor incident is an instantaneous seizure of a reactor coolant pump rotor at 100 percent power with three loops operating.
The tr0nsient was reanalyzed because of the potential effect on the peak reaggor coolant system pressure and fuel temperature. The analysis used +6 X 10 AK/K/'F MTC. The results of the analysis showed that less than 2 percent of the fuel rods experienced departure from nucleate boiling (DNB) and that the peak clad temperature reached was 2250*F. This assures that the i
fuel damage will be minimal, the offsite radioactive release will be a small I
fraction (less than 10 percent) of the 10 CFR 100 guidelines, and that no loss of core cooling capability will result.
The analyses showed that the maximum pressure within the reactor coolant and main steam system did not exceed 110 percent of the design pressures.
Thus is was demonstrated that the analysis results are acceptable.
Loss of External Electrical Load The UFSAR cases analyzed for both beginning and end of life conditions are:
1)
Reactor in manual rod control with operation of the pressurizer spray and the pressurizer power operated relief valves and 2)
Reactor in manual rod control with no credit for pressurizer spray or pressurizer power operated relief valves.
Since the MTC will be negative at and of life, only the beginning of life cases were reanalyzed. The positive MTC will cause increases in both peak nuclear power and pressurizer pressure.
For the first case the reactor trips f - -.
l on high pressurizer pressure. The maximum pressurizer pressure reaches 2520 psia. The minimus DNBR is reached shortly after reactor trip and is greater than 1.30.
For the second case peak pressurizer pressure reaches 2546 psia and the minimum DNBR increases from its initial value throughout the transient.
Since in both cases the DN8 ratio remains well above the 1.3 level and the peak reactor coolant pressure is less than 110 percent of design, the conclusions presented in the UFSAR are still applicable.
Rupture of a Control Rod Drive Mechanism Housina, Control Rod Ejection The rod ejection transient is analyzea at full power and hot standby.
The l
reactivity addition increases nuclear power and hot spot fuel temperatures.
The limiting peak hot spot clad temperature, 2493*F, and the minimum fuel temperature were reached in the hot full power transient. The peak fuel and clad temperatures do not exceed the fuel and clad limits as outlined in the licensee's rod ejection topical (VEP-NFE-2).
Since these criteria i
are more conservative than the requirement of General Design Criterion 28, the results are acceptable.
To evaluate the effect on cperation of NA-1&2 with a slightly positive j
moderator temperature coefficient, a safety analysis of transients sensitive to a zero or positive MTC was performed. This study indicated that the e
i small moderator temperature coefficient does not result in the violation of safety limits for the transients analyzed.
Thus it was concluded that l
the change to a positive MTC will not cause safety limits to be exceeded.
We have reviewed the licensee's submittal and agree with this conclusion.
Therefore, we find the proposed NA-1&2 TS change to be acceptable for a i
full power level of 2775 Wt core power with a maximum reactor coolant system average temperature of 587.8*F.
Environmental Consideration These amendments involve a change in the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20.
The 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 previously published a proposed finding that these amendments involve no significant hazards consideration and there has been no public comment on such finding. Accordingly, these amendments meet the i
eligibility criteria for categorical exclusion set forth in 10 CFR $51.22(c)(9).
Pursuant to 10 CFR 551.22(b), no environmental impact statement or environmental t
assessment need be prepared in connection with the issuance of these amendments.
Conclusion We have 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, and (2) such activities will
g, be conducted in compliance with the Commission's regulations, and 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 Contributor:
M. Chatterton Dated:
December 13, 1985 s
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