Development and Validation of Analysis Method for Reactor Performance and Safety Characteristics of HTGR

Transcription

1 Development and Valdaton of Analyss Method for Reactor Performance and Safety Characterstcs of HTGR Kunyosh TAKAMATSU HTGR Performance & Safety Demonstraton Group Japan Atomc Energy Agency Aprl 2007 Oara, Japan IAEA

2 D&V of software analyss tools for - Reactor knetcs, thermo-hydraulcs, characterstcs of dsturbance of heat utlzaton system, and etc. Reactor Contanment Vessel D&V of software analyss tools for - Behavor of fsson products such as release, transport, and plate-out characterstcs Heat Utlzaton System Vessel coolng system (VCS) 905 IHX : Intermedate heat exchanger PPWC: Prmary pressurzed water cooler SPWC: Secondary pressurzed water cooler IHX SPWC Reactor Auxlary coolng system PPWC Pressurzed water ar-cooler Man coolng system D&V of software analyss tools for - Phenomena of ar ngress, transport of decay heat, performance of coolng system, and etc. D&V of evaluaton methods for - Performance of IHX, Integrty of structure of reactor nternal, hgh-temperature component, and etc.

3 HTGR performance & safety demonstraton group Objectve s to carry out the followng researches usng the HTTR (Hgh Temperature Engneerng Test Reactor) for establshng and upgradng the VHTR (Very Hgh Temperature Reactor) technologes. Enhancement of reactor knetcs evaluaton method Enhancement of nuclear characterstcs evaluaton method Safety demonstraton tests have been conducted to demonstrate nherent safety features wth abnormal status smulaton.

4 Purpose of our research (1/2) Objectve s to develop and valdate an analytcal method of reactor knetcs usng safety demonstraton tests. Reactvty nserton test Partal loss of coolant flow test (one or two out of three gas crculators trp test) Loss of coolant flow test (all gas crculators trp test) Reactvty nserton test Problem s that peak power values of analytcal results are lager than those of the measured values. Reason s that an conventonal method wth one pont knetcs can t consder the temperature dfference (550 deg C) from the nlet to the outlet of the core larger than that of Lght Water Reactor (LWR) and Fast Breeder Reactor (FBR). Method of soluton s to mprove the analytcal model for consderng the dstrbuton of three dmensonal temperature coeffcents.

5 Purpose of our research (2/2) Partal or all loss of coolant flow test Problem s that an analyss code for constructng the HTTR can t analyze the partal or all loss of coolant flow test wth accuracy. Reason s that an analytcal model doesn t contan full core structures and doesn t consder heat transfer from the fuels to the reflector blocks and the RPV. Method of soluton s to mprove the analytcal model for consderng heat transfer from the fuels to the reflector blocks and the RPV.

6 Analytcal model Conventonal model model 1ch 1ch 1TC 1TC Developed model model 4ch 4ch 20TC 20TC Consder the dstrbuton of three dmensonal temperature coeffcents Z Adabatc R Condton Core nlet TC Core outlet ρ = α sngle ΣΔT V ΔT=550 V ρ :reactvty feedback effect Z Reflector blocks R RPV Core nlet Core outlet ρ = Σα ΔT Δ T :Regon temperature rse α :Temperature Coeffcent (TC) V :Volume :Regon number Analytcal method method 1One 1One pont pont reactor reactor knetcs knetcs 2Mult 2Mult flow flow channels 3Regon temperature coeffcents 4Heat 4Heat transfer transfer from from fuels fuels to to Reflector blocks blocks and and RPV RPV k α sngle = Σα = Σ eff 1 k ΔT Temperature rse s small (below 20 deg C) durng the tests. Power dstrbuton doesn't change. eff

7 Reactvty nserton test Permanent reflector block Replaceable reflector block Reactor 原子炉出力 power (%) (%) The 低速 CR 1mwthdrawal m /sec 40m of m 40mm 引抜き高速 5mat m 5mm/sec 40m m 引抜き The CR wthdrawal of 40mm at 1mm/sec Fuel block Center CR Elapsed 時間 tme ( 秒 )(sec) Wthdraw the center control rod The reactor power control system s not operatonal. The other control rods are fxed. Core Reflector block

8 Reactor transent n reactvty nserton test Values of change from the ntal reactor power 初期出力からの変化量 (MW) (MW) ch 1 温度係数 TC 4ch 20 温度係数 TC Measured 実測値 values The reactor power of 15MW The CR wthdrawal of 40mm at 1mm/sec 試験経過時間 Elapsed tme (sec) (sec) The Ths The Ths analytcal fgure fgure shows shows results results the the usng analytcal usng the the 4 results results ch chand usng usng the the 20 the 20 the TC TC 4 ch can ch can and and demonstrate the the TC TC the n the n comparson transents wth of wth of the those the those reactor reactor of of usng usng power power the the better 1 better ch chand than than the the those those 1 TC. TC. usng usng the the 1 ch chand the the 1 TC. TC.

9 Partal or all loss of coolant flow test RPV : Reactor Pressure Vessel Reactor core Trp one, two or three out of three gas crculators PPWC : Measure ponts of coolant flow rate Reactor 原子炉出力 power (%) (%) Gas Crculators PPWC : Prmary Pressurzed Water Cooler Coolant 1 次冷却材流量 flow rate Trp (1 台停止 one GC ) Trp (2 台停止 two GCs ) Reactor 原子炉出力 power Trp (1 台停止 one GC ) Trp (2 台停止 two GCs ) Elapsed 時間 tme (sec)(sec) Results of partal loss of coolant flow tests Coolant 冷却材流量 flow rate (ton/h) (ton/h) Coolant flow rate Reactor power Trp one GC:The reactor power decreased from 60% to 45% Trp two GCs:The reactor power decreased from 60% to 28% Antcpated Transent Wthout Scram Trp one GC:The flow rate decreases from 100% to 66% Trp two GCs:The flow rate decreased from 100% to 33%

10 Reactor transent n partal loss of coolant flow test Reactor 原子炉出力 power (MW) (MW) Reactor 原子炉出力 power (M W (MW) ) Fuel temperature 燃料温度 (Analytcal 解析値 value) Moderator temperature 減速材温度 (Analytcal 解析値 value) Reactor power of 9MW 300 9MW Two 循環機 GCs 2 trp 台停止試験 test Reactor 原子炉出力 power (Measured 実測値 value) Reactor 原子炉出力 power (Analytcal 解析値 value) End 試験終了 of test 試験経過時間 (hr) Elapsed tme (hr) Fuel temperature 燃料温度 (Analytcal 解析値 value) Reactor power of 18MW 18MW 循環機 2 台停止試験 Two GCs trp test Moderator 減速材温度 temperature 解析値 (Analytcal value) Reactor 原子炉出力 power (Measured 実測値 value) Reactor power 原子炉出力 (Analytcal 解析値 value) End 試験終了 of test 試験経過時間 (hr) Elapsed tme (hr) Temperature (deg C) Temperature (deg C) 温度 ( ) 温度 ( ) These From These From the the fgures fgures expermental show show that that the results the results analytcal of of the the results two results two gas gas of of transent crculators reactor reactor trp trp test, test, power power t t are was are was dentcal verfed verfed that to that to the the the the measured reactor reactor power power values values decreases durng durng the due the due tests. to tests. to the the negatve reactvty feedback effect effect Therefore, of of the the core core t t even even s s confrmed f f ATWS ATWS that occurs. that occurs. the the code code s s able able to to smulate the the pre-analyss n n loss loss of of coolant coolant flow flow test. test.

11 Reactor transent n loss of coolant flow test R eactor pow er(%) Elapsed tm e (hr) Ths Although Ths fgure fgure the the shows shows reactor reactor that that power power the the reactor reactor becomes power power crtcal crtcal decreases agan, agan, the the to to decay peak decay peak power power heat heatlevel value value form form s s merely merely the the 2MW. 2MW. maxmum The The reason reason reactor reactor s s that that power power the the core core of of 30MW 30MW temperature due due to to decreases. the the negatve reactvty feedback effect effect of of the the core core even even f f the the all all gas gas crculators trp. trp.

12 Reactvty transent n loss of coolant flow test Reactvty (Δk/k) 5.0E E E E E E E E E E E-03 R-total R-fuel R-mod. R-Xe Elapsed tme (hr) Negatve reactvty s s nserted as as soon soon as as the the all all gas gas crculators trp. trp. After After that, that, the the total total reactvty ncreases due due to to the the decrease of of the the core core temperature.

13 Conclusons The safety demonstraton tests are performed on the HTTR and many valuable data for establshng and upgradng the VHTR technologes have been measured. The mprovement and valdaton of the analytcal model for consderng the dstrbuton of three dmensonal temperature coeffcents and the radal heat transfer of the core s successful to smulate the reactor transent wth accuracy. JAEA wll provde the HTTR data for development of computatonal methods and valdaton for the VHTR system through not only the IAEA CRP but also the GIF VHTR projects.