ML20070S880
| ML20070S880 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 01/26/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070S878 | List: |
| References | |
| NUDOCS 8302080081 | |
| Download: ML20070S880 (4) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 62 TO FACILITY OPERATING LICENSE NO. DPR-66 DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY BEAVER VALLEY POWER STATION, UNIT NO. 1 i
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Introduction By letter dated October 27,1978 (Reference 1) Duquesne Light (the licensee) submitted a proposed amendment to the Technical Specifications for Beaver Valley Power Station. The amendment consists of deleting the requirement for the reactor trip on turbine trip whenever the plant is operated below 50 percent of its rated power. The licensee requested this change because deletion of this trip would lead to an increase in plant availability by reducing the length of time needed to restart a unit following a readily correctable turbine trip at low power. The licensee has provided an analysis indicating that the proposed change would not degrade safe operation of the plant.
I In addition, by letter dated November 3,1982, the licensee shows that the t
probability of a small-break LOCA, resulting from a stuck-open PORV, is substantially unaffected by such modification. This addresses the concern i
of TMI Action Plan Item II.K.3.10. " Proposed Anticipatory Trip Modification."
i Evaluation-I i
1.
Reactor Safety s
The existing Technical Specifications for Beaver Valley Power Station require that the reactor scrams on turbine trip whenever the plant is operating'at a power level higher than 10 percent of its rated power. The change requested by the licensee extends this limit to 50 percent of the rated power. In the analysis provided in support of this change the licensee has shown that when the reactor operating at < 50 percent of its rated power (s1326 MWt) fails to scram on turbine trip, the resultant transient would not cause the plant to exeed its safety j
limits before other reactor protection systems would trip the reactor and bring the plant to the safe shutdown condition.- The licensee has presented the analyses j
for the four cases corresponding to four different sets of assumptions covering the most limiting conditions which could occur after a turbine trip'. In two of i
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them the licensee assumes that the spray systen and the safety relief valves in the pressurizer remain operative and the pressurizer pressure is fully controlled.
In these cases it is. assumed that the reactor scram occurs at 30 seconds after a turbine trip and' is caused by the pump undervoltage resulting from a failure of the network bus transfer which switches the power to the pumps from the i
generator to the external power source. In the other two cases it is assumed l
that the pressurizer spray system and the safety relief valves are inoperative and the reactor scrams on high pressurizer pressure. In both these cases the l
analyses were perfomed with a minimum reactivity feedback in one case and a maximum feedback in the other. The change in reactivity feedback was achieved by postulating zero moderator density coefficient for. the minimum feedback and l
a very large moderator coefficient for the maximisn feedback. Doppler power coefficients were adjusted to provide consistent minimum and maximum reactivity feedbacks. The values of these coefficients are shown below:
l Moderator Density Coefficient:
l 0 to 0.43 ap/g/cc Doppler Power Coefficient:
- 0.84 to -1.57 x 10-4 ap/% power at 52% power In the analysis no credit was taken for steam dump, auxiliary feedwater flow or operation of the safety relief valves in the steam generators. The analysis was performed using threeWestinghouse computer codes: LOFTRAN, FACTRAN and THINC. The LOFTRAN code detemined system parameters, the FACTRAN code was used to calculate the heat flux transient based on the previously detemined nuclear power and the THINC code was employed for computing DNBR values. during the transi ent.
Each analysis was performed for two power levels corresponding to 52 percent and 72 percent of reactor rated power.-
The results of the analysis have incidated that the highest primary system pressure is reached when no credit is taken for the pressurizer pressure control and a minimum reactivity feedback is assumed. The transient is teminated by the reactor scram on high pressurizer pressure before the primary system pressure could exceed its maximum allowable limit.
The lowest value of DNBR occurs when the pressurizer pressure is fully controlled and a minimum reactivity feedback is assumed. In this case a significant decrease in the DNBR is observed, but the DNBR never drops below 1.3 and the reactor scrams l.
safely on ptsnp undervoltage. The above analyses demonstrate therefore that all the pertinent plant parameters remain within their ' safety limits. After i
i reviewing the proposed change to the Technical Specifications we find that the licensee has presented sufficient evidence indicating that whenever a turbine trip occurs in the plant operating at <50 percent of its rated power, it is safe for the reactor to remain operatEg and there is no need for its scram on the signal generated directly by the turbine trip. We conclude therefore that the proposed change of the turbine-trip setting is acceptable from the reactor safety standpoint.
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Plant. System Performance Based on our review of our consultant's Technical Evaluation Report (attached), we agree with his findings that: (1) the licensee analyses i
show that the requested Technical Specification changes do not degrade the other reactor safety sj. ems or change ' their mode of operation; and (2) since the system has the design capability to accept a 50 percent load rejection, a reactor trip is not necessary because of a i
turbine trip until at or above 50 percent power. The proposed license amendment request is,therefore, acceptable from the plant system perfor-mance standpoint.
3.
TMI Action Plan Item II.K.3.10 s
l In a November 3,1982 letter, Duquesne Light states that an analysis was performed by Westinghouse Electric Corporation for bypassing the antici-patory trip on turbine trip below 70 percent power for the Beaver Valley Power Station - Unit 2.
The analysis showed that turbine trip without a direct or immediate reactor trip represents no hazard to the integrity of the reactor coolant system. The analysis used for Unit 2 indicated that, following a turbine trip with the reactor at 70 percent power, the maximum primary system pressure was 2308 psia. With the PORY actuation setpoint at 2350 psia, this results in a 42 psi margin.
The above analysis, performed for Unit 2, used the same assumptions that would apply to an evaluation for Unit I with the exceptior, of the higher power level (70 percent versus 50 percene). Westinghouse Electric Corporation has also indicated that the analysis for Unit 2 is applicable to Unit 1.
Since the analysis for Unit 2 was performed at a higher power level than the 50 percent trip setpoint proposed for Unit 1 and with the same assumptions, we concluded that the Unit 2 analysis envelopes condi-tions which would be applicable to an analysis for Unit 1 with reactor j
trip on turbine trip being bypassed up to 50% power.
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li Based on the results obtained for the Unit 2 analysis and the enveloping aspect of this analysis for Unit 1, we conclude that operation of Beaver V&11ey Power Station - Unit 1 at 50 percent rated power or below with the anticipatory reactor trip on turbine trip bypassed will not signifi-cantly change the probability of a small-break LOCA due to a stuck-open PORV. Therefore, we conclude that the guidance of this TMI Action Plan item is satisfied and the proposed modification can be implemented for Beaver Valley Power Station - Unit 1.
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Environmental Consideration We have determined that the amendment does 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 amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 51.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
f Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated, does not create the possibility of an accident of a type different from any evaluated previously, and does not involve a significant reduction in a margin of safety, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: January 26, 1983 Principal Contributors K. Parczewski P. Shemanski F. Burrows l'
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