ML20024B451

Analysis of TMI-2 Accident W/Continuous Operation of Reactor Coolant Pumps

ML20024B451
Person / Time
Site:	Crane
Issue date:	12/31/1982
From:	GENERAL PUBLIC UTILITIES CORP.
To:
References
TASK-07, TASK-7, TASK-GB GPU-2523, NUDOCS 8307080771
Download: ML20024B451 (22)

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69025k5 ANALYSIS OF THREE MILE ISLAND - UNIT 2 ACCIDENT WITH CONTINUOUS OPERATION OF THE REACTOR COOLANT PUMPS PREPARED BY GPU NUCLEAR CORPORATION 100 INTERPACE PARKWAY PARSIPPANY, N.J. 07054 DECEMBER, 1982 e

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ABSTRACT The Three Mile Island - Unit 2 ('rMI-2) accident that occurred on March 28, 1979 has been analyzed assuming that the Reactor Coolant Pumps (RCPs) were not tripped. The,results show that with the RCPs running throughout the accident, the liquid level in the Pressurizer would have dropped rapidly after closure of the Pressurizer PORV block valve and subsequent operator action based on pressurizer level, using High Pressure Injection (HPI) pumps would have recovered the primary system inventory to a stable subcooled condition. The core would have remained adequately cooled throughout the accident due to the forced circulation of a two phase mixture, and no core damage would have occurred.

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CONTENTS Section Page 1

Introduction 1

2 Analysis of TMI-2 Accident with Continuous 2

Operation of Reactor Coolant Pumps 3

RETRAN Code and TMI-2 Model 14

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4 Conclusions 17 5

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SECTION 1 INTRODUCTION

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This report summarizes the analysis of the effects of continuous RCP operation during the TMI-2 accident of March 28, 1979. - The purpose of this analysis was to investigate the primary system response to the stuck open PORV and throttled RPI with the assumption that the RCP's are maintained in operation throughout. The results show that the liquid level in the pressurizer would have dropped rapidly after closure of the Pressurizer Power Operated Relief Valve (PORV) block valve. Assuming that this would have caused the operators to reactivate the HPI system based on reduced pressurizer level, the primary system would have been refilled to a stable subcooled condition. The core would have remained adequately cooled due to the forced circulation of a two phase mixture, and no core damage would have occurred.

This analysis was performed with the RETRAN (1-1) thermal hydraulic code using a two loop model of TMI-2.

This model was based on previously developed models which have been qualified against actual plant data for a number of applications (1-2, 1-3, 1-4, 1-5).

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SECTION 2 ANALYSIS OF TMI-2 ACCIDENT WITH CONTINUOUS OPERATION OF REACTOR COOLANT PUMPS The TMI-2 Accident of March 28, 1979 has been documented in detail in References 2-1 and 2-2.

The plant response following the loss of feedwater reflected the combined effects of several parameters, including the delay in auxiliary feedwater flow, stuck open PORV, and reduced HPI flow. This analysis performed with the RETRAN 02 MOD 002 code describes the effects of continuous RCP operation during the accident.

l The scenario for the first 74 minutes of the present analysis is the same as the actual accident - a loss of feedwater results in a Turbine Trip followed by a stuck open PORV and reactor trip with auxiliary feedwater delayed for 8 minutes. At 74 minutes and 100 minutes into the actual accident the operator tripped the RCPs in Loop B and Loop A, respectively. However, for l

this analysis, the pumps are assumed to be maintained in continuous opera-tion. The next significant event occurs at 140 minutes when the operator actually closed the PORV Block Valve. Subsequent to this, the operator is l

assumed to control Steam Generator (SG) pressure so as not to allow the plant to heat up.

HPI flow is controlled on the basis of pressurizer level. The analysis shows that leaving the RCPs on results in refilling the i

reactor coolant system to a stable subcooled condition. The sequence of events for this analysis is shown in Table 2-1 and the RETRAN analysis of the above scenario is discussed below.

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The results from RETRAN for the key parameters are presented in Figures 2-1 through 2'-5 with an explanation of the response of key parameters being i

related to the events of the accident for the first 74 minutes since both scenarios are identical for this time period.

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s TABLE 2-1 SEQUENCE OF EVENTS FOR ANALYSIS OF TMI-2 ACCIDENT WITH CONTINUOUS OPERATION OF RCPs Transient Time (sec)

(min:sec)

Event 0.001 0:001 Main Feedwater Pumps Trip 2

0:02 PORV opens 8

0:08 Reactor Trips 480 8:00 Auxiliary Feedwater initiated 8400 140:00 PORV block valve closed and pressurizer level begins to drop rapidly.

9335 155:35 Operator turns on HPI at full flow (Pressurizer level at 120 inches and falling) 10081 168:01 OTSG Secondary isolated 11681 194:41 Operator turns off HPI (Pressurizer level at 200 inches and rising) 12600 210:00 OTSG secondary pressure controlled by operator

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The accident was originated by the nearly simultaneous trips of the main feedwater pumps and the turbine. Both of these occurrences reduced the rate of heat removal from the primary system via the OTSG. The imbalance between heat added to the primary system fluid by the core and that removed via the steam generators resulted in an increasing primary system average tempera-ture and pressure as shown in Figures 2-2 and 2-3.

The increase in pressure caused the PORV to open and the reactor to scram on high pressure af ter 8 seconds. A few seconds af ter the reactor trip, power generation decreased to decay heat power and less heat was being generated than removed via inventory boil-off in the two steam generators. This plus the open PORV caused a reduction in the primary system pressure.

When the primary system pressure decreased to the electromatic relief valve closure setpoint, the valve failed to close and remained stuck in the fully open position, thus continuing to depressurize the primary system. After 30 seconds, the water level in the steam generators reached the setpoint for opening the control valves to initiate auxiliary feedwater (AFW) flow.

However, auxiliary feedwater was delayed for approximately 8 minutes until the operators discovered the AFW isolation valves closed.

Two minutes into the accident, the high pressure injection system came on for 2.5 minutes, which held the primary system temperature at a constant value. At the end of that period pressurizer level was high and increasing' as shown in Figure 2-1 and high pressure injection flow was reduced. The combined effects of the throttled RPI flow and the unavailability of the secondary side heat sink caused the primary system to heat up and at 5.3

r minutes, the temperature in the hot leg reached saturation. Two phase flow continued to escape through the stuck-open PORV upto the time of PORV block valve closure as shown by the high Prassurizer liquid level indication of Figure 2-1.

At about eight minutes the auxiliary feedwater block valves were opened by the operators, and auxiliary feedwater was introduced to the secondary side of the steam generators, which decreased primary system coolant temperature and pressure.

In the actual accident, the operators turned off the 'B' loop RCP's at 74 minutes and the 'A' loop RCP's at 100 minutes. This terminated forced flow core cooling and resulted in subsequent core heatup. By contrast for the present analysis, all four RCP's were allowed to remain in operation which maintained forced flow core cooling, and as a result the predicted system response after PORV block valve closure at 140 minutes is markedly different from the actual accident (2-1).

Figure 2-4 shows a gradually increasing void fraction-for the Loop A cold I

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leg for this simulation, with a maximum void fraction of about 60% being l

achieved, just prior to PORV block valve closure. With pumps running this l

l results in adequate core cooling.

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At 140 minutes, this analysis asaumes the operator closed the PORV Block valve stopping the reactor coolant leakage (this is the same time as in the actual accident). Following block valve closure, the effect of the

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s pressurizer heaters forces liquid out of the pressurizer resulting in a rapid and continual decrease in pressurizer level as shown in Figure 2-1.

As the level decreases to approximately 100 inches, the operator is assumed to turn on HPI at full flow. The injection of cold HPI water into the primary system results in cooling of the primary system which decreases

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primary system temperature and pressure as shown in Figures 2-2 and 2-3.

The continuing injection of HPI water refills the system and results in a decreasing primary system void fraction as shown in Figure 2-4.

The steam generator pressure is shown in Figure 2-5.

As mentioned earlier, auxiliary feedwater flow was delayed eight minutes resulting in the continued depl'etion of water in the steam generators and eventual dryout and termination of heat removal from the primary system by the steam generators for a short period of time. The liquid dryout in the steam generators was reflected by a steady decrease in secondary system steam pressure. The actuation of AFW at eight minutes resulted in a repressurization of the steam generators and the secondary side pressure was then controlled at the turbine bypass system setpoint of 1025 psig.

With RC pumps running, the primary system cooldown with HPI flow af ter PORV block valve closure lowers the secondary system pressure as shown in Figure 2-5.

As the secondary pressure continues to decrease, the operatar is assumed to isolate the steam generators. The primary temperature and pres-sure decrease and secondary pressure decrease continues until all the voids have been collapsed and the pressurizer starts filling up.

At this point,

the RCS is in a refilled and subcooled state, and this causes a re-pressuri-zation and heatup of the primary system and a pressure increase in the secondary system. Subsequent assumed operator actions include the termina-tion of RPI as the pressurizer level increases to 200" and control of the steam generator secondary pressure.

With the plant in a stable subcooled condition, a controlled recovery to cold shutdown conditions would have been carried out.

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TMI-2 Accident with Continuous RCP Operation 500

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TMI-2 Accident with Continuous RCP Operation 1100 PORV blocl< valve closed RETRAN HPI on at 900 full flow 3

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SECTION 3 RETRAN CODE AND 'DtI-2 MODEL Description of RETRAN Computer Program The RETRAN code is the result of an extensive code development effort sponsored by the Electric Power Research Institute (EPRI) since 1975. The objective of the RETRAN project was to develop an improved and reliable thermal-hydraulic program for analysis of light water reactor system transients. RETRAN has now undergone a thorough qualification through the EPRI/ Utility Working Group by comparison of code predictions with experi-D-

mental data, vendor calculations and analytical solutions.

The code can model a Reactor Coolant System (RCS) as an assembly of volumes connected by flow paths or junctions. The volumes specify a region of fluid within a given set of fixed boundaries, whereas junctions represent the common flow areas of connected volumes. A pump or a valve can bo inserted in a flow path. An extensive trip logic mcdel, a non-equilibrium pres-surizer model and a bubble rise model are also included. Heat conductors can represent materials which conduct heat into the fluid of a volume or between the fluids in two different volumes.

Critical flow can be modeled by selecting one of several critical flow options. A point kinetics model 2.s generally used to calculate the normalized power.

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The code uses one-dimensional flow equations to predict the thermal hydraulic transient results. The program solves the mass, energy, momen-tum and optional phase slip equations for subcooled water, two phase steam water mixtures and superheated steam using finite difference methods.

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order polynomials provide the equation of state for steam / water. The one J

dimensional heat conduction equation provides coupling to the fluid energy equation. Energy produced in the core, for example, is removed by the flow of coolant through the core. This is explicitly calculated by the code using basic principles. The application of RETRAN to operational transients J

is especially accommodated through a flexible scheme used to model interaction of control systems with the thermal-hydraulic model.

1 Description of the 1MI-2 Model The RETRAN model used for this thermal-hydraulic analysis consists of 99 volumes, 129 junctions and 27 heat conductors as shown in Figure 3-1.-

The model includes two steam generators, four cold legs and reactor coolant pumps, two hot legs, a vessel downcomer, core inlet plenum, core bypass region and three axial core volumes with three heat conductors. There is also a vessel upper plenum, outlet plenum and a two volume upper vessel head.

In addition, the pressurizer is modeled with its heaters, relief valves and safety valves. The once through steam generators (OTSGs) are each modeled with twelve primary volumes, twelve secondary volumes and twelve heat conductors. The steam generator inlet plenum and outlet plenum are represented as are the downcomers and steamlines. The steamline relief valves are medeled as junctions as is the feedwater input to the steam generators.

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7 SECTION 4 CONCLUSIONS This report demonstrates that had the Reactor Coolant Pumps been kept operational during the TMI-2 accident of March 28, 1979, the core would have remained adequately cooled throughout and the plant would have returned to a subcooled refilled state. This conclusion is based on assumed operator response to pressurizer level using HPI after the predicted pressurizer level collapse following PORV block valve closure.

The validity of the results of this analysis are supported by qualification of EETRAN models of TMI against ectual plant data for a variety of applications.

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SECTION 5 REFERENCES 1-1 "RETRAN - A Program for One-Dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems" Volumes 1-4, EPRI CCM-5, December 1978.

1-2 T. G. Broughton, J. F. Harrison and N. G. Trikouros, "RETRAN, Version 12B Analysis of Rapid Cooldown Transient TMI-2 dual lowered loop, 177 Fuel Assembly 2772 MWe," August 29, 1978.

,1-3 N. G. Trikouros, T.. G. Broughton, J. F. Harrison, " Applications of RETRAN for Three Mile Island Analyses," presented at the ANS Summer meeting in Las Vegas, Nevada, June 1980. ANS Transactions, Volume 34, 1980.

1-4 J. F. Harrison, N. ' G. Trikouros, A. A. Irani, "RETRAN Simulation of the Three Mile Island Unit 1 Turbine Trip Test," ANS Transactions, Volume 32, 1979, Page 455.

1-5 D. J. Denver, J. F. Harrison, N. G. Trikouros, "RETRAN Natural Circulation Analyses dhring the Three Mile Island Unit 2 Accident,"

presented at the ANS Thermal Reactor Safety Topical Meeting, April 8-11, 1980.

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2-1

" Analysis of Three Mile Island - Unit 2 Accident," Nuclear Safety Analysis Center, NSAC-80-1, March 1980.

2-2 "Three Mile Island Unic II Annotated Sequence of Events, March 28, 1979," GPU Nuclear, TDR 044 June 2 1981.

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