ML20236C261

Fort St Vrain Safe Shutdown Using Condensate Sys

ML20236C261
Person / Time
Site:	Fort Saint Vrain
Issue date:	09/28/1987
From:	Ball S, Moses D OAK RIDGE NATIONAL LABORATORY
To:	Heitner K NRC
Shared Package
ML20236C245	List: ML20236C242 ML20236C261
References
CON-FIN-A-9478 TAC-54373, NUDOCS 8710270077
Download: ML20236C261 (8)

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! s TECHNICAL EVALUATION' REPORT.

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FORTST.VRAINNUCLEdhlGENERATINGSTATIONL P

-; DOCKET 50-267

,, . . . . e LICENSEE: PUBLIC. SERVICE CO. 0F COLORAD0' i .-

FORT ST. VRAIN SAFE SHVTDOWN.USING THE CONDENSATE SYSTEM ~

PREPARED BY:

',s S. J. Ball.'

D. L. Moses '

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Oak Ridge National Laboratory. '~

Oak Ridge,' TN.~ 37831~

September 28, 1987

.+ NRC Lead Engineer: :K..L. Heitner' Project: Selected Operating Reactors Issues (FIN A9478), Project 1, Task 8-3' l'

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NOTICE This report was prepared as an account of work sponsored by an agency of the y United States Government. .Neither.the United States Government nor any agency l thereof, or any of their employees, makes any warranty, expressed or implied, or assumed any legal liability or responsibility for any third party's 'use, or the results of such' use, of any information,. apparatus . product or process l disclosed in this report or represents that its use by such third party would not infringe privately owned rights.

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-Technical Evaluation Report Fort St. Vrain Safe Shutdown Using the Condensate System j 1

S. J. Ball D. L. Moses J i

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1. Introduction' II This Technical Evaluation Report (TER) is a follow-up on a previous ORNL TER submitted to NRC on June 8,1987 entitled " Fort St. Vrain Safe Shutdown from 82% Power", by S. J.-Ball and D. L. Moses. This previous TER was included in an NRC letter to PSC dated July 2,1987. It provided independent confirmation of the Public Service.Co. of Colorado (PSC) analyses  ;

for the postulated accident' cases in which only the Environmentally .

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Qualified (EQ) equipment, primarily the firewater system, is.used for cooldown of the FSV reactor following an extended loss of forced circulation (LOFC) event. The reader is referred to this previous TER for background information i and for a description of the ORECA code used in the ORNL calculations.

In this analysis, the cooling capability assumed to be available is via the " Appendix R Condensate Model Train A" (

References:

PSC Letters P-86683,

" Analysis of Firewater Cooldown for 82% Power Operation", Dec. 30, 1986, and P-87055, " Additional Information for Analysis of Firewater Cooldown for 82%

Power Operation", Feb. 17, 1987). This mode uses a revised coolant flow path j scenario in which the first 5 h of cooling is "open loop", utilizing new 6-in seismically-qualified vent lines. Subsequently a closed loop flow path (with reduced flows) is established for the rest'of the cooldown. A schematic of the Train A cooldown system is shown in Fig. 2.1-8 from P-87167 - Report 1,

" Fire Protection Shutdown /Cooldown Model Changes to Appendix R Evaluation".  ;

I The alternative use of Train B as a cooldown path involves other equipment (i.e. the firewater system) that provides greater cooling flows. Hence only the more limiting Train A case is analyzed.

i

2. Accident Scenario Analysis 2.1 Model Parameters and Assumptions The input assumptions for the " reference case" ORECA calculation were similar to those used for the EQ case studies. The important differences were:

1) Long-term operating power (before shutdown) of 82.3%.

2) The updated PSC-supplied estimates of economizer evaporator superheater (EES) cooling water flow to one loop (6 modules) were used per P-87055. For the first 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of the cooldown, there is 700 gpm available at an inlet temperature of 100 F. During this period, the primary helium coolant flow is controlled to prevent the EES outlet water temperature from exceeding 305 F (to prevent boiling). Heat is rejected to atmosphere through the 6-in. vent line. This is referred to as the "open-loop" portion of the cooldown. Subsequently, in the " closed-loop" portion (operating at 1

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higher pressures), the available flow drops to 491 gpm and the water outlet temperature is controlled at 365 F. In this case, the heat is rejected to the decay heat' removal exchanger.

As before, LCS cooling was assumed lost, and there was a'90-min delay in restarting the primary cooling flow.

2.2 Analysis Results i

.As in the previously analyzed EQ results, the ORNL "best estimate" .

predictions of maximum fuel temperatures were lower than those of PSC, while the mean core outlet temperatures were somewhat higher. ORECA's maximum fuel I was 2433 F (1334 C) vs 2875 F (1579 C) per .PSC. Both of these predictions fall below the FSAR maximum. fuel temperature limit of 1600 C. ORECA's maximum average temperature core outlet gas temperature was 1637 F vs 1391 F per PSC.  ;

The calculated circulator inlet temperatures during the cooldowns were well l within acceptable limits.

The ORECA calculations for maximum and average fuel temperature, core' average gas outlet temperature, primary system pressure, and " manually

! controlled" primary helium flow (to prevent EES boiling) are shown in l Figs. 1-3. No flow stagnation (or flow reversals) were predicted to occur during the cooldown. l i

l Sensitivity analyses were done to evaluate margins for error in the ,

models and assumptions; however, the resulting peak temperatures were still generally satisfactory.

In the Proto Power Corporation calculation of the available water ]

inventory for the "open-loop" cooling mode (Attachment to P-87055), a j conservative estimate (without any refilling during the accident) of 3 h is l given. PSC notes in P-87167 that makeup water can be supplied from the main.

cooling water tower basin, and that is presumably sufficient to allow for 5 h of open loop cooling. A case was run in which only 3 h of open-loop cooling was available, and the resulting cooldown was somewhat slower. The maximum predicted fuel and core outlet temperatures were the same as in the reference case, however, since they occurred before the 3 h switchover to closed-loop cooling.  ;

In addition, in a " worst-case" analysis, we assumed that the initial ,

power level is 5% higher, the EES coolant flow is 25% lower, the forced 1 l cooling restar t is delayed 20 additional minutes (to 110 min), and the open-loop coollu; :2de was shortened to 3h. The maximum predicted fuel temperature was 2667 F (1464 C) with a maximum average core outlet temperature of 1672 F.

The sensitivity analysis provided by PSC (P-87158) was also reviewed briefly and found to be in general agreement with ORECA results.

2

2.3 Review of Licensee's Condensate Flow Calculations A brief review was made of the thermal-hydraulic simulation models developed by Proto-Power Corporation and applied to the analysis of condensate flow in the Appendix R Train A analyses. . As concluded in the previous TER on EQ safe shutdown cooling, the methods (Attachment 6 to P-87055) and models (Appendix A of Attachment 2 to P-87055) appear.to be reasonable; however, there appear to have been no alternate calculations performed to verify the Train A condensate flow analyses. There were no documented applications of the Train A flow calculational model to analyses of normal decay heat removal at Fort St. Vrain using condensate. Presumably, numerous occasions have existed where known levels of decay heat were removed by condensate supply through the emergency condensate line to one steam generator EES and one helium circulator using one 12-1/2 percent-condensate pump. Although such a closed loop configuration would not duplicate the initial condition for Train A open loop operation, a benchmark comparision of. the analytical models to measured normal shutdown cooling conditions would provide higher confidence in the analytical technique. As documented in Region IV Inspections and '

Enforcement Report 50-267/87-14, the NRC audit of the licensee's . independent verification checks of the contractor's analyses for safe shutdown cooling appears to have addressed Proto Power calculation 82-12 for the Appendix R Train A analyses. The inspection report does not indicate specifically that these calculations were reviewed. It is recommended that NRC confirm that the previous audit did address the expected verification check of the Train A analyses.

3. Conclusions Results of the " Appendix R" scenario analyses indicate that there is substantial margin for operational error, equipment degradation, and conservative calculational assumptions before significant fuel damage or fission product release would be predicted. As before, with the EQ case, it is assumed that suitable procedures and operator training are in place such that the emergency systems can be operated properly. Of special note is the need to manually control the primary coolant flow such that steaming and choking are prevented in the EES. However, the changes which occur in the operating parameters are so slow that ample time is allowed for the operators to make appropriate adjustments in the circulator speed.

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