ML20235F515

Safety Evaluation Re Safe Emergency Shutdown of Reactor Sys. Operation at 82% Acceptable

ML20235F515
Person / Time
Site:	Fort Saint Vrain
Issue date:	07/02/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20234C612	List: ML20234C609 ML20235F508 ML20235F515 ML20235F528 ML20235F544
References
TAC-63576, NUDOCS 8707130423
Download: ML20235F515 (4)

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Enclosure 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO SAFE EMERGENCY SHUT 00WNS (REACTOR SYSTEMS)

PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION DOCKET NO. 50-267

1. 0 INTRODUCTION In Fort St. Vrain License Event Report (LER) #86-026, dated October 17, 1986, the Public Service Company of Colorado (PSC) reported that the Safe Shutdown

' Cooling System for removing the decay heat following the postulated " Design Basis Earthquake" or " Maximum Tornado" accidents was inadequate. PSC stated in this LER that if one of these two accidents were to occur while the reactor was operating at 105% power, and if, as postulated in Section 10.3.9 of the FSAR, the functions of all non-seismic, non-Category 1 components were lost and the primary helium coolant flow was assumed interrupted for 90 minutes to allow for manual realignments, the safe shutdown cooling system would be unable to keep the fuel temperature below the 2900' F limit.

Further this LER states that the analysis for the removal of decay heat by the Safe Shutdown Cooling System, "did not consider firewater pump capacity nor the associated steam generator inlet or discharge piping configurations." j For the corrective action in the LER, the PSC committed to reanalyzing this Safe Shutdown Cooling System and providing an acceptable method to remove the decay heat and cool the plant without fuel temperatures exceeding 2900*F. ]

In the two-loop Fort St. Vrain plant each loop has six steam generator modules which have parallel secondary coolant flow paths. Each steam generator has two sections, i.e., an economizer-evaporator-superheater '

(EES) section and a reheater section.

The reheater sections of the steam generators are much smaller than the EES l sections; so their use seemed logical for removing the smaller decay heat load. However, the reheaters are designed for steam, not water, so their 4 cross sectional flow area is relatively high. The consequence of this is that the firewater pumps have only enough flow capacity to flood one or two reheater sections, rather than all six as previously assumed. PSC's re-analysis showed that this partial flooding would not provide enough heat transfer area; so PSC concluded that the reheater sections should not be used for the Safe Shutdown Cooling.

Instead, PSC proposed to use the EES sections by initially venting them to the atmosphere. However, the available vent path was not redundant or seismically qualified, so new 6 inch vent lines had to be installed. Even 8707130423 870702 7 DR ADOCK 050

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with these new vent lines PSC found that the fuel temperature could not be l kept below the 2900 F limit for the Safe Shutdown Cooling accident scenario from 105% reactor power.

PSC had analyses done to determine from what power level safe, emergency shutdowns could be accomplished for all of these accident scenarios. These analyses, which PSC submitted to the NRC, showed that, depending on the accident scenario, the fuel temperature can be kept below the 2900 F limit during emergency shutdowns after long-term operation at power levels up to and including 82 percent power.

2. 0 EVALUATION l l

The NRC had the Oak Ridge National Laboratory evaluate all of these sub-mittals. The technical evaluation report (TER) on this evaluation is Enclosure 4. Three parts of the TER pertain to this Safety Evaluation (SE).

The fourth part, which is on possible structural and metallurgical failures in the steam generators, is the subject of a separate safety evaluation.

The NRC staff has reviewed the ORNL TER and agrees with ORNL's evaluations l and conclusions, except as addressed below.

l The first of the three parts is the evaluation of the calculations of the maximum fuel temperature that will be obtained after these postulated accidents. This evaluation was made by using the Oak Ridge developed ORECA

! computer program to independently calculate these temperatures. As can be i

seen in the TER the ORECA calculations show that 82% is a conservative power l level for a limiting fuel temperature of 2900 F. We concur with this finding in the ORNL TER that the 82% power limit proposed by the licensee l is acceptable.

The second of the three parts in the ORNL TER that pertain to this SE is the evaluation of the ability of the existing systems to supply sufficient water flow to both the helium circulator pelton wheel drives and the EES sections of the steam generators during these emergency cooldowns. The final con-clusion of this lengthy review is that for these scenarios, "there is substantial margin in the existing cooling systems to provide-for a safe shutdown." This conclusion is contingent on several items, two of which with we concur and restate as follows:

1. There are operating procedures for these accidents and that the operators have been trained to follow them.

2. PSC should perform an explicit analysis to demonstrate that the original Class I firewater flow path can accommodate a single active failure in the new Class I firewater flow path when the required EES pre-cooling times and the long term cooling are accounted for.

Another contingency in the ORNL TER findings is for the NRC to perform an audit or do confirmatory analyses of the PSC flow calculations. However, the staff believes that with the satisfactory agreement between PSC's cal-culated results and the results of the firewater flow test, which are m

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y reported in Reference 2 and mentioned on page 7 of the TER, no confirmatory I analyses are required. (However, the staff has requested NRC Region IV to (

perform an audit of the licensee's independent verification of these  ;

i calculations.) l The remaining contingency in the TER findings and conclusions is concerned with a passive failure after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of cooling. However, PSC's calcu-I lations show that after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of cooling adequate flow can be obtained from a redundant flow path. Based upon these calculations, the staff finds that a passive failure can be accommodated.

t By letter dated June 24, 1987, the licensee has stated that: l (1) operating procedures have been provided for the postulated " Design Basis Earthquake," " Maximum Tornado," and " Appendix R Fire" accidents and the operators are trained to follow them; and (2) all of the redundant firewater flow paths can accommodate a single active failure up to 83.2 percent power. (This includes EES pre-cooling times and long-term cooling.)

On this basis we conclude that the first, second, and fourth conditions described above provide an acceptable basis to satisfying the requirements of the second part of the ORNL TER, and that the licensee has shown that the existing systems can supply sufficient water during emergency cooldowns.

The third of the three parts in the TER that pertain to this SE is the evaluation of the possibility of water hammer that would prevent these emergency cooldowns. The ORNL TER agrees with the licensee's con-clusions that a water hammer is unlikely because of the steam generator design, and water hammer forces would be reduced by the restriction of the tube entrances. The staff further notes that steam generator modules are designed for an inlet pressure of about 3180 psia (Table 4.2-7 of the FSV FSAR). By contrast, the firewater pumps have a total design head of only 140 psia (Section 9.12.3.3 of the FSV FSAR).

It is difficult to conceive how the low pressure output of the pump can cause damage to a system designed for over 20 times that pressure.

Hence, the staff concludes that it is highly unlikely that a water hammer )

will preclude a safe shutdown.  ;

3.0 CONCLUSION

S The staff finds that the Fort St. Vrain reactor can be shutdown after prolonged operation at 82 percent of the licensed power without having the fuel temperature exceed the 2900* F limit. Thus the staff finds that operation at 82 percent power is acceptable.

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4.0 REFERENCES

1. SECY-77-439 dated August 17, 1979.

2. Letter from H. L. Brey, Public Service Company of Colorado, to J. A. Calvo, USNRC, dated May 4, 1987.

Principal contributor: E. Lantz, RSB Dated: July 2,1987 1

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