ML20207F057

Safety Evaluation Supporting Util 870206 Submittal Re Safe Shutdowns During Postulated Accident Conditions

ML20207F057
Person / Time
Site:	Fort Saint Vrain
Issue date:	08/10/1988
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20207E958	List: ML20207E953 ML20207F057
References
TAC-66574, NUDOCS 8808180207
Download: ML20207F057 (5)

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%, . . . . . jl SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO SAFE SHUTDOWNS OURING POSTULATED ACCIDENT CONDITIONS l l

PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN REACTOR DOCKET NO. 50-267 l

1.0 INTRODUCTION

I In October 1986 (Ref. 3), the Public Service Company of Colorado (PSC) reported the inability of Fort St. Vrain (FSV) to provide emergency safe shutdown i capabilities from 105 percent licensed power under certain conditions. The ,

licensee concluded that the previous analysis did not consider the capacity of 1 the alternate pump nor the complete piping configuration of the shutdown l cooling flowpath. PSC followed up with analyses including those additional considerations to detemine the maximum power level from which FSV could  !

achieve emergency safe shutdown cooling. This value was detemined to be 87.5 '

percent power. The Appendix R shutdown scenario was also reanalyzed and found to limit the FSV power level to 82 percent, bounding the safe shutdown cooling .

scenario. The staff found the analyses acceptable and approved operation for l FSV for up to 82 percent power (Ref. 1), j However, the questionable adequacy of the Safe Shutdown Cooling System gave ,

rise to a question of the adequacy of those shutdown cooling assumptions used in accident analyses. Therefore, the licensee submitted reanalyses via Reference 2 for those postulated Chapter 14 accidents dependant upon the safe shutdown cooling system for fuel failure protection. The reanalyses are the i subject of this review. l l

2.0 EVALUATION j

2.1 Background

The Safe Shutdown Cooling System serves to provide alternative decay heat l removal capabilities during shutdown conditions. Decay heat removal normally I takes place in the two Fort St. Vrain (FSV) steam generator (SG) regions -- the economizer-evaporator-superheater (EES) section and the reheater section.

Secondary water flows thru the SG tubes; primary coolant (helium) flows over i the outside of the SG tubes.

Shutdown cooling water is supplied as both a source of heat sink coolant and a source of circulator motive power. One or more of three shutdown coolant sources are used: firewater, condensate, and feedwater. The coolant flow  ;

directed to the SG provides the heat sink coolant, while flow to the Pelton l wheel powers the helium circulators providing the helium flow through the core p

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necessary for shutdown cooling. A single circulator has the capacity needed to conduct shutdown cooling following all accidents except a design-basis primary l

depressurization accident.

For those postulated accidents analyzed in Chapter 14 of the FSV FSAR, those considered to be most limiting with respect to shutdown cooling ability are those events involving excessive heat removal due to an increase in primary I heat production, a loss of feedwater supply, or a loss of secondary inventory, l Rapid primary depressurization, loss of normal shutdown cooling, total interruption of forced cooling, and secondary-to-primary leakage are included asthegostlimitingtypesofaccidents. A minimum heat removal rate of I 73.5x10 8tu/hr is required for shutdown cooling following power operation at .

83.2% power. l 2.2 Accident Reanalyses Of the eleven types of accidents analyzed in the FSAR Chapter 14, three include '

postulated accidents that depend upon safe shutdown. cooling to protect fuel integrity (See Table 1). These are loss of Nomal Shutdown Cooling, Design ,

Basis Rapid Depressurization, and Secondary Coolant System Leakage.

l During a loss of Nomal Shutdown Cooling, the secondary coolant source may be either firewater, feedwater, or condensate. Onefirewsterpumpwithaboosger pump has sufficient flow capacity to provide a heat removal rate of 86.7x10 Btu /hr with 20'F subcoo'ing maintained in the SG. One 121% condensate pump at lower capacity, bug hisher pressure than the firewater pump, provides heat ,

removal of 73.6x10 Btu /hr with 15'F subcooling maintained in the SG. The '

nomal capacity of a single feedwater pump as used during shutdown cooling is capable of complete heat rer: oval during normal operation, and is rore than adequate for heat removal during shutdown cooling. The interruption of forced cooling (10FC) for 30 minutes was reanalyzed for the condensate, firewater, and feedwater flew paths discussed above, and was determined to provide shutdcwn i adequate cooling for the duration in each case. Each of the above Loss of  !

Nomal Shutdown Cooling Accidents was found to supply adequate shutdown cooling fron 83.2% power.

The motor-driven feedwater pump is the assumed source of secondary coolant for )

the Rapid Depressurization Accident. With the high feedwater pump capacity. l greater than either the firewater or condensate pump, a much higher flow rate, l and consequently a6 much higher degree of subcooling, could be maintained. Heat removal at 76.4x10 Btu /hr results and 216'F subcooling is maintained in the SG during shutdown cooling from 105% sower. The limiting condition for this  ;

cooldown scenerio is the rate at witch the low density primary coolant can 1 transfer heat frmn the core to the SG. '

Of the Secondary Coolant System Leakage events postulated for FSV, the bounding '

event for shutdown cooling capability is the "Subheader Rupture and Moisture Detector Failure. The severity of the accident depends upon the resultant ingress of secondary coolant water into the primary coolant system. Water

3 reacts with the graphite in the core by oxidation, depleting the graphite shell of the fuel particles and also generating hydrogen gas and carbon monoxide.

Radioactive )roducts, such as Cesium 137, can be released by this process.

Failure of t1e moisture detector with the rupture of a SG subheader was found to yield the highest secondary-to-primary leak.

Reanalyses were performed for the limiting Moisture Detector Failure case and for the Wrong loop Dump case for subheader rupture. Each case was analyzed I assuming condensate cooling by each SG section -- EES and reheater sections --

independently and the worse case was found to be the Moisture Detector Failure.

The reactor is conservatively assumed in this event to trip on the high PCRV pressure trip instead of the high PCRV moisture trip, delaying the assutred reactor trip and increasing the total assumed leakage. The results of the analyses showed the reheater case to be most limiting.

The worst case analysis resulted in about a 20,000 lbm ingress of water / steam into the primary system. A maximum of 0.17'4 of the graphite in the core was depleted, which is bounded by the 51 fuel coating failures assumed in the  !

design activity bases. Acccrding to the analysis, the heat up rate exceeded the cooldown rate for two hcurs. However, the peak graphite temperature rerained within the 1200'F graphite-water reaction threshold. A gracual cooldown followed. The combined heat removal by the SG reheater and by the FCRV liner cooling system provided enough cooling to prevent excessive primary pressures and temperatures, maintaining adequate shutdown cooling from 100%

power. ,

1 2.3 Computer Codes 1

, In their analyses for heat removal capacit leak, the Transient Analysis Program (TAP)y was following a secondary-to-primary used. A steady state code called SUPERHEAT verified the results. Hand calculations were done to verify 1 TAP heat balance results, as well. The TAP code has been referenced by the l licensee and approved by the NRC staff for use in previous licensing actions, and is acceptable in this application.

Analyses to determine heat transfer capabilities were done using FRDRNEW and 4

FSVSGNEW computer codes, developed under NRC sponsorship. The heat removal rates were calculated with FSVSGNEW using system pressure drops generated by )

1 PRDRNEW for the three shutdown cooling paths. The codes were developed by Proto Power Corporation and are acceptable as used in this application.

3.0 CONCLUSION

S We have reviewed the reanalyses of those FSAR Chapter 14 accident analyses which assume cooling by the safe shutdown cooling system. Based on the results previded by the licensee, we conclude that the shutdown cooling system is adequate to provide at least the minimum cooling capacity as required to ,

prevent fuel terperatures from exceeding 2900'F using additional flow path l

considerations (firewater pump capacity and actual secondary ficw paths and l

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l theirassociatedpipingconfigurations)notusedinthepreviousanalysesfor Chapter 14 of the FSAR. +

4.0 REFERENCES

1. Letter from T. E. Murley to R. O. Williams, dated July 2, 1987.  ;

2. LetterfromH.J.Brey,PublicServiceCompanyofColorado(PSCo)toH.N.

Berkow, USNRC, dated February 6, 1987.

3. Fort St. Vrain LER #86-026. October 17, 1986. l 4 Letter from H. L. Brey, PSC, to J. A. Calvo, USNRC, dated May 4,1987.

Principal Contributor: A. P. Gilbert, SRXB Dated: August 10, 1988 I

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TABLE 1 CHAPTER 14 ACCIDENT ANALYSES WITH l 5AFE SHUTDOWN COOLING l

l l 1. Loss of Normal Shutdown Cooling (a) Cooling with One Water-Turbine Driven Circulator Driven by Feedwater (b) Cooling with Dne Water-Turbine Driven Circulator Driven by Unboosted Condensate or Boosted Firewater (c) Cooling with One Feedwater Driven Circulator During Loss of Helium Pressure at "Maximum Credible" Rate (d) Total Interruption of Coolant Flow for Thirty Minutes

2. Design Basis Rapid Depressurization/Blowdows

3. Secondary Coolant System Leakage (a) Steam Generator Leakage into the PCRV (Prestressed Concrete ReactorYessel)--SubheaderRuptureandWrongLoopDump (b) Steam Generatcr Leakage into the PCRV --Subheader Rupture and Moisture Detector Failure i

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