ML20041C263
| ML20041C263 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 02/04/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17341A908 | List: |
| References | |
| NUDOCS 8202260424 | |
| Download: ML20041C263 (4) | |
Text
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?g UNITED STATES NUCLEAR REGULATORY COMMISSION y
g WASHINGTON. D. C. 20555 3*
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 76 TO FACILITY OPERATING LICENSE NO. DPR -31 AND AMENDMENT NO. 70 TO FACILITY OPERATING LICENSE NO. DPR-41 FLORIDA POWER AND LIGHT COMPANY TURKEY POINT PLANT UNIT NOS. 3 AND 4 DOCKET NOS. 50-250 AND 50-251 INTRODUCTION By letter dated December 10, 1981, as supplemented on January 20 and 28,1982, Florida Power and Light Company (the licensee) requested amendments to Facility License Nos. DPR-31 and DPR-41 for the Turkey Point Plant Unit Nos.
3 and 4.
The licensee requested a change to the Turkey Point Units 3 and 4 moderator temperature coefficient (MTC) Technical Specification. The request proposes to increase the upper bound of the MTC from 0 to 0.5 X 10-5 ag/g*F for power l evels below 70". of rated power.
EVALUATION The licensee provided a safety analysis that assessed the impact of a positive moderator temperature coefficient of reactivity (MTC) on the accident analyses presented in Chapter 14 of the Turkey Point Units 3 and 4 Final Safety Analyses Report (FSAR). Those transients which were found to be sensitive to a positive or near zero MTC were reanalyzed. These are limited to transients which cause the reactor coolant system (RCS) temperature to increase. Transients that result in a reduction in RCS temperature for which a negative MTC is more limiting; and those for which heatup effects following reactor trip are not sensitive to MTC were not reanalyzed.
f The transients not reanalyzed are:
I:
A.
RCCA misalignment / drop 4
B.
Startup of an inactive RCS loop C.
Excessive heat removal due to feedwater system malfunctions D.
Excessive load increase E.
Loss of normal feedwater, loss of offsite power F.
Rupture of a main steam pipe G.
Loss of coolant accident (LOCA) i 8202260424 820204 PDR ADOCK 05000250 l
P PDR
7-5.
The transients reanalyzed, with.one exception, used a MTC=+5 x 10-5 agfgfop assumef to remain constant for variations in temperature. The exception is the rod e s_ tion accident for which the model assumes the MTC becomes less positivo a
at higher temperatures. This is acceptable since the MTC is actually zero or negative above 70% power as required by the proposed Technical Specification.
The transients analyzed and their results are:
A.
Boron Dilution The reactivity addition due to a boron dilution at power will cause an increase in power and RCS temperature. Due to the temperature increase a positive MTC would add additional reactivity and increase the severity of the transient. With the reactor in automatic control, the rod insertion alarms provide the operator with adequate time to terminate the dilution before shutdown margin is lost. With the reactor in manual control the baron dilution incident is no more severe than a rod withdrawal at power, which is analyzed in item C below.
B.
Control Rod Withdrawal from a Subcritical Condition This transient results in an uncontrolled addition of reactivity leading to a power excursion causing a heatup of the moderator and fuel. The time the core is critical before a reactor trip is very short so that the RCS temperature does not ingrease significantly; hence the effect of a positive MTC is small. The analysis results show a transient average heat flux which does not exceed the steady state full power value and an increased core water temperature that remains below the full power value.
To provide assurance that the above cirteria are sufficient bases for acceptance, the licensee submitted comparison analyses (letter from R. Ulrig, FPL to D. Eisenhut, NRC, dated Janaury 28, 1982) with reactors that use newer approved methods to calculate DNBR using power distributions that would occur during the transient. These results show that the DNBR remains above 1.3 during the transient. This is acceptable since it is expected that the Turkey Point reactor would have similar results for the rod withdrawal error from a subcritical condition.
C.
Uncontrolled Control and Assembly Withdrawal at Power This transient produces a mismatch in steam flow and core pmver, resulting in an increase in RCS temperature. However, the results show that the i
nuclear flux and overtemperature AT trips prevent the core minimum DNS ratio from falling below 1.3 for this transient, so the conclusions presented in the FSAR are still valid.
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9 D.
Loss of Coolant Flow The most severe loss of flow transient is caused by the simultaneous loss
'i of power to all three reactor coolant pumps (RCP's). This case was reanalyzed to determine the effect of a positive MTC on the nuclear l
power transient and the resultant effect on the minimum DNBR reached a
during the transient. The RCS temperature increases 6*F above the i
initial value and a minimum DNBR of 1.6 is obtained for this transient.
Since this is the limiting loss of flow transient present in the FSAR l
and since the DNBR ratio remains above 1.3 the results from the FSAR are still valid and acceptable.
E.
Locked Rotor The locked rotor event was reanalyzed because of the potential effect of f
the positive MTC on the nuclear power transient and thus on RCS pressure
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and fuel temperature. A positive MTC will not affect the time to DNB i
because ONB is conservatively assumed to occur at the beginning of the transient. The results show that the FSAR analysis at 100% power and a O MTC is more limiting than the +5x10-5 aK/g/*F MTC at 70% power. The i
peak fuel pellet average ternperature reached during transient was 2137'F, the peak cladding temperature was 1587'F, and the peak RCS pressure was 2430 psia, which do not exceed the accepted safety limits as presented in the FSAR.
F.
Loss of External Electric Load j
The loss of external electric load transient was reanalyzed for beginning f
of cycle (BOC) since the MTC will be negative at end of cycle (EOC) and l
will give the same results as in the FSAR. Two cases were analyzed.
(1) reactor in the automatic rod control mode with operation of the pressurizer spray and pressurizer power operated relief valves (PORV);
and (2) reactor in the manual control mode with no credit for pressurizer i
i spray or PORV's. The result of a loss of load is a core power that momentarily exceeds the secondary system power removal, causing an increase in RCS coolant temperature. The reactivity addition due to t
7 a positive MTC, causes an increase in both nuclear power and RCS The result for the control rods in the automatic control pressure.
and assuming pressurizer spray and relief is an RCS pressure of 2443 psia following a reactor trip on overtemperature AT. A minimum DNBR l
, of 1.74 is reached shortly after reactor trip, The result for the case i
of rods in manual control with no credit for pressure control is a peak RCS pressure of 2534 psia following a reactor trip on high pressure.
The minimum 01BR is initially 1.86 and increases throughout the transient.
Since the Of:8 ratio remains above 1.3 and the peak RCS pressure is less than 110% of design the conclusions presented in the FSAR are still e
applicable.
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. G.
Control Rod Ejection The rod ejection transient was analyzed only for BOC since the MTC will be negative at EOC and the previous analysis (Unit 4 Cycle 4 Reload Safety Evaluation) results are applicable. The high nuclear power levels and i
hotspot fuel temperatures resulting from a rod ejection are increased by a positive MTC. The results of BOC reanalysis show that the fuel and L
clad temperatures are within the limiting values specified in the FSAR and the Unit 4, Cycle a Reload Safety Evaluation and as such are acceptable.
The peak hotspot fuel centerline temperature exceeded the melting f
temperature for the full power case; however, melting was restricted to less than the innermost ten percent of the pellet. The maximum fuel enthalpy reached 177 cal /gr which is below the specified limit of 200 4
cal /gr stated in the Cycle 4 Reload Safety Evaluation.
SUMMARY
i Since the reanalysis of the affected plant transients do not result in exceeding any of the fuel limits or safety limits specified in the Turkey l
Point Units 3 and 4 FSAR or Cycle 4 Reload Safety Evaluation, we conclude that the proposed Technical Specification will not pose an undue risk to the health and safety of the public, and is therefore acceptable.
ENVIRONMENTAL CONSIDERATION We have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendments involve an action which is i
insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 151.5(d)(4), that an environmental impact statement or negative i
declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
i CONCLUSION We have concluded, based on the considerations discussed above, that: (1) because the amendments to not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered t
t by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
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Date: February 4,1982 j
Principal Contributors:
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R. F. Frahm H. Richings l
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