ML20151B665
| ML20151B665 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 04/01/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20151B663 | List: |
| References | |
| TAC-54373, NUDOCS 8804110106 | |
| Download: ML20151B665 (3) | |
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ENCLOSURE 1 SUPPLEMENTAL SAFETY L/ALUATION BY THE OFFICE OF huCLEAR REACTOR REGULATION CONCERNING APPENDIX R SAFE SHUTDOWN DECAY HEAT REMOVAL CAPACITY PUBLIC SERVICE COPPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION _
DOCKET N0. 50-267
1.0 INTRODUCTION
This'Supplenental Safety Evaluatian (SSE) is concerned with the compliance of the Fort St. Vrain Nuclear Generating Station (FSV) with 10 CFR Part 50, Appendix R concerning fire prote tion for nuclear power facilities.
Specifically, it discusses the shility of the decay heat removal system to assure that the no fuel failure occurs following postulated fires in the non-congested cable areas.
This SSE is required because the staff's overall Safety Evaluation (SE) en corpliance with Appendix R, dated December 18, 1987 identified this as an open issue (See Section 2.1.?.3 a) 2) of the above SE). The licensee has proposed means to shutdown and cool the reactor consisting of two trains (A and B) of post-fire safe shutdewn equipment which provide for i
reactivity control, PCRV integrity, and decay heat removal.
As part of this review, the staff questioned the effectiveness of the flow paths through the steam generators to remove reactor decay heat. The supporting analyses were submitted by PSC letters dated December 30, 1986 (P-86683), and May 1, 1967 (P-87159). Based on a review of the available information, the cor.ceptual designs of the flow paths are accept.able provided:
(a)theaboveanalysisverifiestheeffectiveness of the flow path, and (b) sufficient makeup water capability is demon-strated. The review of these analysis will be the subject of this SSE.
2.0 EVALUATION 2.1 Train B - Firerater Tra g l
For this evaluation, the first Appendix R safe shutdown train discussed is the firewater train.
In this train, the diesel driven fire water pump provides flow from the main cooling tower through a steam generator and a helium circulator. These flows are vented to the atmosphere, or returned to the turbine building surp. The system operates open loop. This train is essentially the same as the Safe Shutdown Cooling system already evaluated by the staff and its 4
8804110106 880401 PDR ADOCK 05000267 F
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- contractor, Oak Ridge National Laboratory (ORNL). These evaluations are contained in a letter dated July 2, 1987. The conclusions of this prior evaluation are that safe shutdown can be accomplished from 82 percent of full power without fuel damage.
Thus, on the basis of the staff's prior evaluation acceptance of this train as having adequate decay heat removal capacity, we conclude this train is also acceptabic for Appendix R safe shutdown.
2.2 Train A - Cendensate Train The second Appendix R safe shutdown train is the condensate train.
In this train, condensate pump IC provides condensate flow from the decay heat rer. oval (DHR) Exchanger through a steam generator and a heliun circulator and back through the DHR Exchanger. The effective-ness of this train in rereving decay heat is the subject of this evaluation. The staff requested its contractor, ORNL to perform an independent evelvation of the condensate train, and its decay heat
' removal ca pacity. Enclosure 2 is ORNL's Technical Evaluation Report cencerning this ratter. CRNL evaluated the decay heat renoval via the condensate train, using the same models that were previously used to evaluate the firewater train. These models and their results were previously accepted by the. staff in a SE dated July 2, 1987. The NRC l
staff has reviewed tho ORNL TER and agrees with ORNL's evaluations and conclusions, except as addressed below.
The first part is the evaluation of the calculations of the maxirium fuel ter:perature that will be obtained after these postulated cooldown scenario. This evaluation was made by using the Oak Ridge developed ORECA c:mputer program to independently calculate these temperatures.
As can be seen in the TER the ORECA calculations show that 82% is a conservative power level for a limiting fuel temperature of 2900'F.
The maximum fuel temperature is nearly 500 F below this limit, indicating the system has a wide margin of safety. We concur with this finding in the ORNL TER that the 82% power limit proposed by the licensee is acceptable.
The second parts of the ORNL TER is the evaluation of the ability of the existing systems to supply sufficient water flow to both the heliun circulator pelton wheel drives and at the steam generators during this cooldown scenario. -The ORNL TER concludes that the analysis.rethods "appear reasonable", but recorcends that the staff reconfirm a previous audit of this area which included the Train A and B calculations. However, the staff notes that the firewater pump delivers a ninimum flow rate of about 800 gpm against a back pressure of no more than 100 psia, as verified in the earlier analysis of the Train B perfortnance. A single ccrdensate pump delivers a minimum flow rate of 670 gpm at 300 psi or 800 gpm at
reduced pressure. Hence, we conclude that because of the hicher back pressure on the condensate pump, and because of the corparable effec-tive flew rates of both trains relative to pump back pressures, adequate flow can be developed by the condensate pump for the Train A cooldown scenario.
The staff also notes that ORNL's own sensitivity aralysis shews that even if the flows are less than expected, there is still a wide temperature margin to fuel failure.
Hence, the staff concludes the condensate train has adequate flow for decay heat reroval.
3.0 CONCLUSION
S l
The staff and its contractor, ORNL have reviewed the licensee's calculations j
for decay heat removal under two Appendix R scenarios. The staff concludes 1
that Fort St. Vrain Appendix R shutdown systers Train A and Train B have adequate decay heat removal capacity.
Date: April 1, 1988 Principal Contributor:
K. Heitner i
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1 TECHNICAL EVALUATION REPORT FORT ST. VRAIN NUCLEAR GENERATING STATION DOCKET 50-267 LICENSEE:
PUBLIC SERVICE CO. OF COLORADO o
FORT ST. VRAIN SAFE SHUTDOWN USING THE CONDENSATE SYSTD4 PREPARED BY:
S. J. Ball D. L. Moses Oak Ridge National Laboratory Oak Ridge, TN. 37831 September 28, 1987 FRC Lead Engineer:
X. L. Heitner Project: Selected Operating Reactors Issues (FIN A9478), Project 1, Task 8-3 (Y '7 / r n ~ -,. </ 91 v/tvl/Gui a
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i NOTICE This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency 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 s
disclosed in this report or represents that its use by such third party would not infringe privately owned rights.
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Technical Eva uation Report Fort St. Vrain Safe Shutdown _Using the Condensate System i
S. J. Ball D. L. Moses 1.
Introduction 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 j
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 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 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 "Appbndix 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 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".
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.
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 rpm 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
higher pressures), the available flow Jrops 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 As in the previously nnalyzed 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 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 within acceptable limits.
The ORECA calculations for maximum and average fuel temperature, core average gas outlet temperature, primary system pressure, and "manually controlfed" primary helium flow (to prevent EES boiling) are shown in Figs. 1-3.
No flow stagnation (or flow reversals) were predicted to occur during the cooldown.
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 conservative estimate (without any refilling during the accident) of 3 h is given. PSC notes in P-87167 that makeup water can be supplied froe 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 terperatures 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 254 lower, the forced cooling restart is delayed 20 additional minutes (to 110 min), and the open-loop cooling mode 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 (T-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 reasoneble; 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 Trmin A flow calculational model to analyses of normal decay heat removal at For', 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 j
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 adjustuents in the circulator speed.
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