SAFETY DEMONSTRATION TESTS ON HTR-10

Transcription

1 2nd International Topical Meeting on HIGH TEMPERATURE REACTOR TECHNOLOGY Beijing, CHINA,, September 22-24, 24 #Paper H6 SAFETY DEMONSTRATION TESTS ON HTR-1 Shouyin HU, Ruipian WANG, Zuying GAO Institute of Nuclear Energy Technology, Tsinghua University P.O. Box 121,Beijing 1221, China ABSTRACT: The 1MW High Temperature Gas-cooled reactor (HTR-1) is a graphite-moderated helium gas-cooled reactor with the thermal power of 1MW and the reactor outlet temperature of 7. Safety demonstration tests on the 1MW High Temperature Gas-cooled Reactor (HTR-1) were conducted to demonstrate the inherent safety features of HTR-1 as well as to obtain the core and primary cooling system transient data for validation of safety analysis codes. Two simulated anticipated transients without scram (ATWS) tests, lose of heat sink by tripping the helium blower without scram and reactivity abnormal increase by means of withdraw a control rod without scram were selected as first phase safety demonstration experimental items. This paper describes the tests with detail experimental method, experimental condition, measures for guaranteed safety and experimental results. The test of tripping the helium circulator ATWS was conducted on October 15,23. Helium blower was switched off during normal operation with 3MW, that means primary coolant system was disconnected. Although none of 1 control rods was moved, the reactor power immediately decreased due to the negative temperature coefficient. After 5 minutes, the reactor becomes criticality again. The output power went to a stable level with about 2kw. The test of reactivity increase ATWS was carried out two times. The quantity of reactivity increased were Δk/k and Δk/k by a control rod withdrew 12cm and 13cm with moving speed 1cm/s respectively during reactor normal operation with 3MW. Following the control rod withdrew, the reactor power increased, the maximum power level reached to 537kW and 723kW corresponding reactivity of Δk/k and Δk/k. Then, the engineering safety features executed safety actions, helium circulator stopped, primary and secondary circuits insolated. The reactor power decreased sharply by means of negative reactivity feedback due to the increased of core temperature, and reached to a stable level about 2kW after 2 hours later. The main parameters such as helium pressure, graphite structure s temperature, metallic support s temperature and pressure vessel s temperature is far below the safety limitation during both tests conducted. KEYWORD: HTR-1, Gas-cooled reactor, Safety features, Inherent safety, Lose of heat sink, Reactivity abnormal increase, ATWS

2 Safety Demonstration Tests on HTR-1 #H6. INTRODUCTION The 1MW High Temperature Gas-cooled reactor (HTR-1) is a graphite-moderated helium gas-cooled reactor with the thermal power of 1MW and the reactor outlet temperature of 7ºC. Safety demonstration tests on the HTR-1 were conducted to demonstrate the inherent safety features of HTR-1 as well as to obtain the core and primary cooling system transient data for validation of safety analysis codes. HTR-1 is the first modular high temperature gas cooled reactor designed and constructed by institute of nuclear and new energy technology of Tsinghua University of China. HTR-1 reached its first criticality in December 2 and reached its rated parameter in January 23. The safety demonstration tests were conducted after fulfilled all the planed commissioning test items and obtained the basic performance of systems and components of HTR-1. Two simulated anticipated transients without scram (ATWS) tests, lose of heat sink by tripping the helium blower without scram and reactivity abnormal increase by means of withdraw a control rod without scram were selected as first phase safety demonstration experimental items. The tests were conducted at 3% rated power conditions. The test of lose of heat sink ATWS was conducted on October 15,23. The test of reactivity increase ATWS was carried out two times, reactivity addition is Δk/k and Δk/k, respectively. 1. BASIC FEATURES AND PRE-TEST ANALYSIS 1.1 Basic features related to test The commissioning results, such as reactivity worth of control rod, reactivity temperature coefficient, ability of cavity cooling system, temperature distributing in core during normal operation and shutdown, are basic data and preconditions for carried out the safety test. (1) Reactivity worth of Control rod The reactivity worth of control rod was calibrated with period method at helium atmosphere. the total reactivity worth of each control rod is Δk/k, and the differential reactivity worth- is shown in Fig.1 [1]. (2) Temperature reactivity coefficient The temperature reactivity coefficient of HTR-1 is Δk/k/ºC that measured by means of the reactivity equilibrium at different average temperature of core and reflector[2]. The helium circulator and no more than 5kW nuclear power were used to heat the reactor together, and the inlet temperature is near to 25ºC, normal operating condition, finally. (3) Coasting of helium circulator The rotate speed transient during coasting of helium circulator tripped at 3% rated state is shown in Fig.2. The flow rate of helium is in proportion to circulator s rev, so it varies in the same way as Fig.2. (4) The power dissipated by surface cooling system: 26kW[3] (5) There is 4kg water in the steam generator. (6) The neutron kinetics parameters (calculated): Effective delayed neutron fraction β=.726 Prompt neutron lifetime λ=.168s (7) Calculated temperature coefficient [4] coefficient of fuel temperature: Δk/k/ºC coefficient of moderator temperature: Δk/k/ºC coefficient of reflector temperature: Δk/k/ºC 1.2 Pre-analysis

3 HTR24 Beijing, CHINA, Before the experiments, pre-analysis about the tests had been processed using the HTR-1 safety design and analysis code THERMIX in order to validate the test conditions and to estimate the aftereffect The ATWS test on the trip of the helium circulator (1) The conditions of test the reactor power: 3kW the pressure of primary circuit: 25kPa the temperature of outlet helium: 25ºC the temperature of inlet helium: 65ºC the time of closing the baffle of primary circuit: 15seconds the time of isolating the secondary loop: 9seconds (2) The hypothesis events sequence for analyzing, is listed in Table 1 Table 2 the hypothesis events sequence after tripping helium circulator Event Tripe of helium circulator 18 Protection system initiated by signal of ratio of water flow to helium flow is high. The baffle of primary circuit stars closing. 19 The secondary loop starts isolating. (3) The analyzing results The flow rate of coolant in primary circuit decreases rapidly after stopped the helium circulator. The reactor power, transient behavior is shown in Fig.3, is decreased due to negative reactivity feedback caused by the rising of temperature in the core because of the decreasing of heat removal capability of primary circuit. After about 36 seconds, the reactor power is reduced to the level of decay heat. During An hour, the maximum pressure in the primary circuit is 285kPa; and the maximum temperature of fuel elements is about 823ºC, which is far below the safety limitation of 123ºC The ATWS test on withdraw of one control rod (1) The conditions of test the reactor power: 3kW the pressure of primary circuit: 25kPa and 3kPa the temperature of outlet helium: 25ºC the temperature of inlet helium: 65ºC the time of closing the baffle of primary circuit: 15seconds the time of isolating the secondary loop: 9seconds the control rod withdrawal rate: 1mm/s the positive reactivity addition: Δk/k and Δk/k (2) The hypothesis events sequence for analyzing, is listed in Table 2 Table 2 the hypothesis events sequence after withdraw of one control rod Event Withdrawing one control rod 32.4 Protection system initiated by signal of power is high

4 Safety Demonstration Tests on HTR-1 #H Helium circulator is tripped. The baffle of primary circuit starts closing. The secondary loop starts isolating (3) The analyzing results Two cases, reactivity addition is Δk/k and Δk/k,are calculated. Transient behavior of reactor power is shown in Fig.4. The reactor power is increased with inducing the reactivity by withdraw the control rod. When the reactor nuclear power rose up to setpoint, the protection system is initiated. The helium circulator is switched off, and the primary loop and the second loop are isolated. But, because the control rods are not inserted, the reactor power is increased unceasingly till the reactivity due to negative reactivity feedback caused by the temperature rising in the core is larger than the reactivity adding by withdrawal of control rod. Then the reactor power is decreased rapidly to the level of decay heat, and the reactor is in subcritical. After about 4 minutes, the reactor becomes criticality again caused by the temperature decreasing in the core. The maximum nuclear power is 499kW and 418kW, the maximum temperature of the fuel elements is about 855ºC and 882ºC corresponding to reactivity addition of Δk/k and Δk/k, respectively, which is far below the safety limits.. During the process, the pressure in the primary loop does not rise up obviously. 2. BYPASS DEVICE OF SHUTDOWN BREAKER FOR ATWS 2.1 Why need a bypass device When tripping of the helium circulator, the protection system could be initiated by shutdown signals of helium flow rate is low and the ratio of water flow to helium flow is high. When withdrawing one control rod continually, the protection system could be initiated by shutdown signals of the increase rate of reactor power is high and reactor power is high. To maintain the control rods a position during the tests, the bypass device of shutdown breaker is needed to avoid lose of the control rods supply. Then, the tests can be conducted as an ATWS. 2.2 Function requirement For safety and preventing the wrong operation by an operator, the bypass device should meet the following requirements: (1) Bypassing the rector scram function only (2) Don t affect the functions of protection system (3) Don t affect the scram by manual mode (4) Don t affect the working of control rods (5) Having clear indication of running state (6) Can t be put into bypass mode before protection system running normally 3. TEST METHOD 3.1 Parameters measured in the tests The main parameters to be measured by a distributed computer system in the tests are as follows: (1) Reactor power (2) Inlet and outlet temperatures of the core (3) Reflector temperature (4) Pressure of primary circuit (5) Temperatures of pressure vessel

5 HTR24 Beijing, CHINA, (6) Helium flow rate 3.2 Limitations of safety related parameters The limitations of operating parameters approved by National Nuclear Safety Authority (NNSA) to insure the systems and components of HTR-1 in safety state during test are listed as follows: (1) Reactor power: <=12kW (2) Temperature of pressure vessel: <=35ºC (3) Temperature of metal supports in PV: <=4ºC (4) Pressure of primary circuit: <=33kPa, the relief valve don t open (5) Temperature of shielding: <=7ºC 3.3 Test procedure The test procedure is shown in Fig.5. First of all, the initial conditions and special measures to cope with abnormal must be confirmed. Then, the scram breaker is bypassed. Performed above jobs, the helium circulator is stopped, or one control rod is withdrawn, and then the transient behaviors begin. When achieved test purpose or gone beyond limit of any parameter, the reactor is shutdown by manual scram mode. 4 TEST RESULTS 4.1 The ATWS test results on the trip of helium circulator Initial test conditions When HTR-1 in normal operation with power of 3kW,outlet helium temperature of 65ºC, inlet helium temperature of 215ºC and pressure of primary loop of 25kPa,the helium circulator was switched off after bypassed the reactor scram breaker Test results Main events sequence is shown in Table 3.The transients of the reactor power and rotate speed of helium circulator, inlet and outlet helium temperature, the pressure of primary circuit, the temperature of components in core, temperature of pressure vessel are shown in Fig.6 to Fig.1. After the helium circulator was switched off, the reactor power was decreased due to negative reactivity feedback caused by the rising of temperature in the core because of the decreasing of heat removal capability of primary circuit. After about 7 seconds, the reactor power is reduced to the level of decay heat. The variation of core temperature caused the reactor became critical again after 26 seconds. Experienced several oscillations, the reactor power went to stabilization at 2kW that was dissipated without any problem by the surface cooling system after about 72 seconds. The temperature of top reflector was increased significantly 162ºC in two hours, and the temperature of side reflector upper side was increased slightly also, but the temperatures of other parts of core structure were decreased slowly. The temperatures of pressure vessels were not increased except the top cover of RPV. The pressure of primary circuit was decreased slowly [5]. Table 3 the test results of tripping helium circulator Events Helium circulator was switched off

6 Safety Demonstration Tests on HTR-1 #H6 12 Protection system was initiated by signal of ratio of water flow to helium flow is high. The baffle of primary circuit was closed. The secondary loop was isolated The control rods didn t insert 7 Power reduced to decay heat 252 Reactor became critical again 72 Power went to stabilization at 2kW 4.2 The ATWS test results on withdrawal of control rod Initial test conditions When HTR-1 in normal operation with power of 3kW,outlet helium temperature of 65ºC,inlet helium temperature of 215ºC and pressure of primary loop of 25kPa, one control rod was withdrawn at normal moving speed from a position where to top the value of reactivity changed is Δk/k after bypassed the reactor scram breaker. The maximum change rate of reactivity was Δk/k/s Test results This test was conducted two times, in the first time the reactivity addition was Δ k/k,and in the second time the reactivity addition was Δk/k.. Main events sequence is shown in Table 4.The transients of the reactor power and rotate speed of helium circulator, inlet and outlet helium temperature, the pressure of primary circuit, the temperature of components in core, temperature of pressure vessel are shown in Fig.11 to Fig.14. Table 4 the test results of withdraw one control rod Events Withdrawing one control rod 12 Protection system was initiated by signal of power increasing rate is high. The baffle of primary circuit was closed. The secondary loop was isolated 35 Peak power of 723kW 81 Power reduced to decay heat 24 Reactor became critical again 72 Power went to stabilization at 2kW In the second time, the reactor power was increased rapidly with withdrawal of the control rod. The protection system was initiated by signal of power increasing rate is high after 12 seconds. The helium circulator is switched off, and the primary loop and the second loop are isolated. But, because the control rods was withdrawn till to top, the reactor power was increased unceasingly till the reactivity due to negative reactivity feedback caused by the temperature rising in the core is larger than the reactivity adding by withdrawal of control rod. The maximum nuclear power reached to 723kW. Then the reactor power is decreased rapidly to the level of decay heat, and the reactor is in subcritical. After about 24 seconds, the reactor became criticality again. Experienced several oscillations, the reactor power was stabilized at a very low power level. Compared the test results of two times, it can be inferred that the peak power is high, the time became critical again is early with the reactivity addition is large

7 HTR24 Beijing, CHINA, The temperature of top reflector was increased significantly 215ºC in two hours, and the temperature of side reflector upper side was increased slightly also, but the temperatures of other parts of core structure were decreased slowly. The temperatures of pressure vessels were not increased except the top cover of RPV. The pressure of primary circuit was decreased slowly [6]. 5. EVALUATION OF THE TESTS RESULTS 5.1 the factors of affecting tests results The trend of transients of reactor power and temperature between pre-analysis and tests results is agreement, but the value of power and the time of reactor becoming critical again have some difference. The transient of pressure of primary circuit experiences a rising phase in pre-analysis, but it is decreased slowly in tests. The main reasons caused the difference are (1) using a conservative value of temperature coefficient and don t calculate the xenon in pre-analysis;(2) the affection of decay heat; (3) a part of helium in the primary circuit is not circulated. (1) The affection of decay heat The variation of decay heat after reactor scram is difference with the normal operation time. The decay heat will be larger if the operation time is longer. The decay heat cannot affect the transients in first short time of ATWS test, but it can affect the time and power level of reactor becoming critical again after auto shutdown. If the decay heat is lower, the reactor will become critical again in a shorter time due to the temperature of core decreasing faster. (2) The affection of temperature coefficient and xenon The rate of reactor power decreasing in pre-analysis is lower than in tests, because of the value of temperature coefficient using in pre-analysis is lower than measured. Meanwhile, because the xenon isn t taken into account in calculating, the time is lower, but power level is higher than them in test when the reactor becomes critical again after auto shutdown. (3) The variation of pressure Only the circulated helium can cause the variation of pressure of primary circuit in calculate. For compensating the affection of static helium, the circulating space in analysis model is enlarged. Considering this part of static helium is gathered in the space of low temperature, this space is located at the upper of inlet in pre-analysis model. When the helium circulator is switched off, the flow direction of hot helium changes from downward to upward, so the temperature of helium at upper of core increased rapidly. The helium space on the core is small in realistically, so attached a cold helium space at the inlet is not suitable. 5.2 Modifications of calculate model and final analysis According the tests results and above mentioned analysis in 6.1, the modifications of calculate model are as following: (1) The temperature coefficient is addition 5% of pre-analysis (2) The volume compensated the static cold helium is placed at the bottom of RPV. Using the renewed calculate model, the transients of tests are analyzed again. The transients of reactor power and pressure of primary circuit in ATWS of trip of helium circulator are shown as Fig.15 And Fig.16. Because the xenon isn t taken into account in new model still, the time is lower, but power level is higher than them in test when the reactor becomes critical again after auto shutdown. The variation of pressure is agreement very well between analysis and test

8 Safety Demonstration Tests on HTR-1 #H6 In analyzing the transients of ATWS of withdrawal of control rod, except adopting renewed model, the control rod is withdrawn from middle position, however, the control rod is withdrawn from bottom position in pre-analysis. That means the rate of reactivity addition is bigger than in pre-analysis, but it is same as in test. The results are listed in Table.5.The calculated results are agreement with the test results. Table 5.calculation results on withdrawal of control rod (renewed model) Items Test Calculate The time of protection system was initiated (s) The time of reached to peak power (s) The value of maximum power (kw) The first time of reactor reached to critical (s) CONCLUSION It is the first time conducted such kind tests in modular high temperature gas cooled reactor. The results obtained from these tests confirmed the safety features of HTR-1 practically. Without any of control rod inserted, the negative temperature coefficient of the reactivity can shutdown the reactor automatically and bring the reactor power safely to a stable low level in the case of helium circulator trip. During the tests, the heat generated in the core was passively dissipated by a surface cooling system without causing unacceptably high temperature in the components. The boundary of primary circuit was integrated without additional active gas released. REFERENCE [1] Shouyin HU, commissioning report, calibration of reactivity worth of control rod of HTR-1, INET, 22(in Chinese) [2] Shouyin HU, commissioning report, measurement of the temperature coefficient of HTR-1, INET, 22(in Chinese) [3] Xihua LIANG, commissioning report, measurement of the ability of surface cooling system of HTR-1, INET, 23 (in Chinese) [4] Final safety analysis report on HTR-1, INET, 1998 (in Chinese) [5] Shouyin HU, commissioning report, ATWS test of helium circulate trip of HTR-1, INET, 23(in Chinese) [6] Shouyin HU, commissioning report, ATWS test of withdrawal of control rod of HTR-1, INET, 23(in Chinese)

9 HTR24 Beijing, CHINA, mk mm Fig.1 control rod worth measured Rotate speed (%) Fig.2 rotate speed after helium circulator trip

10 Safety Demonstration Tests on HTR-1 #H Power (kw) Power (kw) Power 2.5mk Power 5mk Time (s) Fig.3 transient of power in helium circulator trip (pre-analysis) Fig.4 transient of power in withdraw control rod (pre-analysis)

11 HTR24 Beijing, CHINA, Start Confirmation of initial conditions and special measures Bypass the scram breaker Stop helium circulator or Withdraw control rod Measurement of required parameters >Limit? Yes No No Achieve? Yes Manual shutdown of reactor End Fig.5 test procedure Power(kw) power(kw) speed(rpm) Fig.6 transient of power after helium circulator trip (test) Speed (rpm)

12 Safety Demonstration Tests on HTR-1 #H6 Temperature ( ) Time Fig.7 transients of inlet and outlet temperatures after helium circulator trip (test) Temperature ( ) top reflector 2 side reflector upside 3 side reflector underside 4 outlet mixing room 5 bottom reflector 6 metal support Time (s) Fig.8 transient of core temperature after helium circulator trip (test)

13 HTR24 Beijing, CHINA, hot gas duct Temperature ( ) 15 1 RPV top RPV bottom RPV body Time (s) Fig.9 transient of RPV temperature after helium circulator trip (test) 3 Helium pressure (kpa Time Fig.1 transient of pressure after helium circulator trip (test)

14 Safety Demonstration Tests on HTR-1 #H6 Power(kw) power(1mk) power(5mk) position(5mk) position(1mk) Rod poaition (mm Fig.11 behavior (short) of power in withdrawal of control rod (test) Power (kw) position(5mk) position(1mk) power(5mk) power(1mk) Tme (s) Fig.12 behavior (long) of power in withdrawal of control rod (test) Position (mm

15 HTR24 Beijing, CHINA, Temperature( ) outlet mixing room side reflector underside bottom reflector bottom reflector metal support side reflector upside Fig.14 behavior of core temperature in withdrawal of control rod (test) Power (1%) Calculate Test Time (s) Fig.15 transient of power after helium circulator trip (calculation and test)

16 Safety Demonstration Tests on HTR-1 #H6 Pressure (bar) Pre-analysis Final analysis Test Time (s) Fig.16 transient of pressure after helium circulator trip (calculation and test)