ML19296B866
| ML19296B866 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 02/08/1980 |
| From: | Baynard P FLORIDA POWER CORP. |
| To: | Ross D NRC - TMI-2 UNRESOLVED SAFETY ISSUES TASK FORCE |
| References | |
| 3--3-A-3, 3-0-3-A-3, NUDOCS 8002220281 | |
| Download: ML19296B866 (19) | |
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N,4.f Florida Power L O H PO N A f tO N February 8, 1980 File:
3-0-3-a-3 Mr. D. F. Ross, J r.
Director Bulletins & Orders Task Force Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Identification and Resolution of Long-Term Generic Issues Related to the Commission Orders of May 1979
Dear Mr. Ross:
Enclosed is Florida Power Corporation's response to Item 1A of Attach-ment A to your August 21, 1979 letter.
This response provides a generic benchmark analysis of sequential aux-iliary feedwater flow to the OTSG's following a loss of main feedwater.
This analysis used the CRAFT-2 code with the 3 node steam generator model used in small break evaluation and demonstrates the ability of this model to calculate steam generator heat removal effects with rea-sonable accuracy.
If you require further discussion concerning our response, please con-tact this office.
Very truly yours, FLORIDA POWER CORPORATION Y
P.
Baynard Manager Nuclear Support Services PYBemhR04D70 g
8002220 Attachment General Office 3201 Thirty-fourtn street soutn. P O Box 14o42. st Petilhg. Florida 33733 e 813-866-5151 l
STATE OF FLORIDA COUNTY OF PINELLAS P. Y. Baynard states that she is the Manager, Nuclear Support Services of Florida Power Corporation, that she is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto, and that all such statements made and matters set forth therein are true and correct to the best of her knowledge, information and belief.
er:w iv. F. Baynded Subscribed and sworn to before me, a Notary Public in and for the State and County above named, this 8th day of February,1980.
0ma u a) n Notary Public Notary Public, State of Florida at Large, My Commission Expires: August 8, 1983 CameronNotary 3(D12)
so.
Response to Question 1A Provide a benchmark analysis of sequential auxiliary feedwater flow to the steam generators following a loss of main feedwater. This analysis was provided in a letter from J. Taylor (B&W) to R. Mattson (NRC) dated June 15,1979. However, in this analysis the TRAP-2 code with a 6 node steam generator model was utilized. All small break analyses presented to the NRC have been performed using the CRAFT-2 code with a 3 node steam generator model. We require a benchmark analysis for sequential auxiliary feedwater flow also be performed using CRAFT-2 with a 3 node steam generator representation.
by The Babcock & Wilcox Company Nuclear Power Generation Division 4
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C0rlTENTS
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'IIITRODUCTIO!!
II SITE EVENT DESCRIPTION III METHODS IV RESULTS Y
CONCLUSIDflS FIGURE 1 CRAFT-2 N0DItiG DIAGRAM FOR SMALL BREfKS FIGURE 2 STARTUP FEEDWATER FLOW STEAMGENERdTORLIQUIDLEVEL(TEMPERATUREADJUSTED)
FIGURE 3 FIGURE 4 STEAM GENERATOR SECONDARY SIDE PRESSURE FIGURE 5 PRIMARY A LOOP TEMPERATURE FIGURE 6 PRIttARY B LOOP TEMPERATURE FIGURE 7 PRESSURIZER LEVEL FIGURE 8 REACTOR VESSEL PRESSURE 4
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1.
INTRODUCTION This report presents an analysis of sequential auxiliary feedwater (AFW) flow to the once through steam generators for a loss of main feedwater transient. The CRAFT 2 code l and the small break model described in reference 2 have been used in the study.
The calculated results have been compared to a loss of offsite power startup test data obtained from the Florida Power Corporation's Crystal River 3 Unit in which an imbalance in the auxiliary feedwater flows between the two operating loops resulted in an imbalance in the primary loop response.
This transient tests several features of the computer simulation, including conditions of asymmetric loop temperatures, an almost dry generator to feed auxiliary feed-water into, loss of RC pumps, and establishment of natural circulation.
In many cases the absolute validity of the boundary conditions and test data were ques-tionable, and estimates had to be used. However, this analysis does show that the data trends can be predicted by a 3 node CRAFT 2 SG representation.
II.
SITE EVENT DESCRIPTION The Crystal River 3 Unit is a 2452 HNt, 177-FA B&W reactor with a lowered-loop configuration.
On April 23, 1977, a loss of offsite power test was performed.
This test was initiated from approximately 15% full power operation.
The secon-dary liquid levels were approximately 2 feet and was sufficient to remove the power and provide essentially steady-state operation prior to test initiation.
The test was initiated by tripping the reactor, the reactor coolant pump, and feedwater pump power sources.
The core power then dropped to the decay heat level and, as the primary coolant pumps coasted down, the primary flow decayed to natural circulation level. One diesel generator was started to provide power for the pressurizer heaters, one makeup pump, and other necessary services of secon-dary importance to this analysis.
The main feedwater flow coasted down resulting in both steam generators eventually drying out until the auxiliary feedwater flow became sufficient to start filling the A loop steam generator secondary at about two minutes into the transient.
The B loop steam generator remained dryed out until twelve to fourteen minutes into the transient when the A loop reached normal operating level and the feed-water flow was diverted to the B loop.
The imbalance in the feedwater flows, and hence levels, resulted in a corresponding imbalance in the primary system re-sponse including the decay' heat removal, the hot and cold leg temperatures and 54 e
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The transient results were used to evaluate the ability of the 3 node CRAFT 2 steam generator model used in small break evalua-tions to calculate the effect of the feedwater transient.
III. METHODS
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A.
CRAFT Input Model The input model developed for this calculation was based on the small break model used for licensing.2 The schematic of the flow path nodalization is shown in Figure 1.
The initial system conditions were defined based on the available men-sured data which were requir2d to represent this test.
The model was set up to Provide a steady-state calculation until two seconds into the transient when the reactor, reactor coolant pumps, and main feedwater pumps were tripped initiating the transient calculation.
3.
Initial Conditions The initial mass flow was assumed to be identical to the full power operation value. The measured hot and cold leg temperatures were then used to determine a consistent core power to provide the initial steady-state operating conditions.
This resulted in an initial power of 19% of full power operation versus the 15%
power defined in the summary test report.
Hand calculations, using the 15% core power and the measured hot and cold temperatures, resulted in a mass flow con-siderably below that required to balance the pump power.
The actual mass flow is believed to have been only 1 or 2% less than full power flow.
The pressure dis-tribution around the system was revised, because of the new hot and cold leg temperatures, to maintain the loss coefficients defined by the referenced model.
The liquid levels in the pressurizer and steam generator secondary were changed to reflect the measured data.
C.
Boundary Conditions The makeup pump flow was modeled by defining the pressure flow characteristic curve for normal operation with the recirculation line open.
The makeup pump was actuated when the pressurizer level dropped to 30" below the initial liquid level value. The makeup pump flow was equally distributed between the two cold leg pump discharge modes as shown in Figure 1.
The feedwater flows were defined by the test data and are given in Figure 2.
An auxiliary feedwater entnalpy of 58 btu /lbm, which is the nominal enthalpy of the system,was used.
The safety relief valves were set to 1030 psia to model the effect of the turbine bypass valves, which are fully open at 1030 psia. The safety relief flow is the only allowance made in the model for steam flow.
The heat transfer to the secondary was assumed to be to the mixture in the lower portion of the steam generator and the fraction which may have been deposited in tne steam region was assumed to be negligible. A preliminary short-term transient evaluation demonstrated the need to define the heat transfer multiplier based on the steam generator secondary levels. Consequently, the final model contained a heat transfer multiplier as a function of tLme based on the measured secondary levels.
IV.
RESULTS This section presents a comparison of the CRAFT 2 analysis to the data taken for the first 20 minutes of the CR-3 loss of offeite power test. As will be shown, some of the data utilized in the evaluation is questionable and greatly influence the transient response. However, even with the uncertainties in the measured data, the CRAFT 2 code is shown to adequately calculate the RCS behavior.
A.
Secondary Response Figure 3 shows the secondary side SG 1evels during the test.
The test data shows that, following the loss of main feedwater, the initial IcVel in both staan gen-erators decreases. At approximately 1 minute into the transient, the auxiliary feedwater system initiates, as shown in Figure 2, and preferentially feeds the A loop steam generator. Thus, the liquid level in SG A increases. At 12 minutes, the liquid level in SG A stabilizes because it has reached its control point. At that time, the feedwater flow is diverted to SC B and its level increases.
The CRAFT 2 code calculated results shows reasonable agreement with the SG A level during the first 12 minutes. After this time, however, the CRAFT 2 calculation continues to increase the SG 1evel while the data shows a level stabilization after this time.
This difference is probably due to an overestimation of the auxiliary feedwater flow to SG A after this time. The auxiliary feedwater flow, as indicated in Figure 2, is very stable and at a relatively high flowrate after 12 minutes. Examining other data, such as the A loop hot and cold leg tempera-tures, does not support a high auxiliary feedwater flowrate.
In light of the ability of the CRAFT 2 code to reasonably predict the SG response up to 12 min-utes and the inferences obtained from other data, the flowrate given in Figure 2 after 12 minutes is believed to be in error.
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The SG B liquid level response is generally overpredicted by the CRAFT calcula-tion. This again is believed to be caused by an overestimation of the auxiliary feedwater flowrate to SG B, especially between 3 and 9 minutes.
Figure 2 shows the auxiliary feedwater flow to be very low over this time period and very stable.
This may be due to an initial instrumentation offset and no feedwater may have been delivered to the steam generator in this period. Once a sustained auxiliary feedwater flow is established to the SG, the CRAFT, calculated level increases are in reasonable agreement with the data.
Figure 4 shows the SG secondary side pressure response during the transient.
CRAFI2 predicts the pressure response for t' a A loop SG reasonably. Between 4 and 6 minutes, the calculated SG pressure increases above the data. Over this time period, it is believed that the measured auxiliary feedwater flows are low.
This conclusion is consistent with the level comparison shown in Figure 3.
For the remainder of the transient, the prediction is higher than the measured SG
, pressure.
The secondary side pressure for SG B was generally underestimated throughout the transient. This is caused by condensation of the steam within the SG due to the excess auxiliary feedwater flow utilized in the calculation.
B.
Primary System Response Figure 5 shows the A loop temperature response during the test. The hot leg tem-perature compares well with the transient data until 13 rJ.nutes. After this time, the CRAFT 2 calculation' continues to show a decrease in the hot leg temperature due to the continued feeding of the A loop SG.
The data shows a flattening of the hot leg temperature due to the control of the SG level. This supports the belief that the auxiliary feedwater flows after 12 minutes is lower than the values indicated by Figure 2.
The calculated A loop cold leg temperature response is consistent with the data trend, but generally overpredicts the data after 4 minutes.
This is caused by the overprediction of the SG A secondary pressure discussed previously.
The B loop temperature response is shown in Figure 6.
Due to the overprediction in the B loop SG level and underprediction in the SG pressure, the hot leg tem-peratures are underpredicted.
Figures 7 and 8 show the pressurizer level and system pressure comparison.
Hand calculations which were performed indicate that these parameters are not consis-tent. Examining these figures, it is seen that the calculated pressurizer level response is in good agreement out to approximately 12 minutes. After 12 minutes, the continued overcooling of the A loop, due to the overestimation of feedwater flow, results in an underestimation of the pressurizer level.
The pressure response shown in Figure 8 shows that the CRAFT 2 calculation under-predicts the data. However, as mentioned previously, this is not unexpected as the system pressure and pressurizer level are not consistent.
V.
CONCLUSION A sequential auxiliary feedwater flow transient has been benchmarked in this analysis using the CRAFT 2 code with the 3 node SG model used in small break evaluations. The site data trends were reasonably reproduced by the code.
In many cases the validity of test boundary conditions were questionable and esti-mates of the test data were used. However, the results provide assurance that the CRAFT 2 code is capable of reasonably predicting the primary system behavior indicated by the test if the boundary conditions were well defined.
Thus, this study has demonstrated that, in spite of the simplicity of the CRAFT 2 steam generator model, the CRAFT 2 code can estinate, with reasonable accuracy, a tran-sient highly dependent on the steam generator.
Thus, the ability of the small break model to calculate the effect of steam generator heat removal during a small break transient is reasonably assured.
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REFERENC_E_S, S
3.R.A. Hedrick, J.J. Cudlin, and R.C. Foltz, " CRAFT 2 Fortran Program for Digital Simulation of a Multinode Reactor Plant During Loss-of-Coolant,"
BAW-20092, Rev. 2, Babcock & Wilcox, April 1975.
2 Letter J.H. Taylor (B&W) to S.A. Varga (NRC), July 18, 1978.
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