Start-up and Safety Analyses of Super Fast Reactors
,2/2NJ>*H5)83E=
$ H?L
Sutanto
,-0.
Start-up and Safety Analyses of Super Fast Reactors
,2/2NJ>*H5)83E=
$ H?L
! E4 7H4 G8,6;BB? B9 7I4 A687 ,6<8A684 A7 A: <A88E<A: 0 4 F874 . A<I8EF<GL 4 A7 ! E4 7H4 G8,6;BB? B9 A: <A88E<A: -B>LB <GL. A<I8EF<GL
BBC8E4 G<I8' 4 =BE<A( H6?84 E A8E: L + 8F84 E6; BA + 84 6GBE- ;8BEL
Sutanto
,-0.
Start-up and Safety Analyses of Super Fast Reactors
Abstract
Start-up and safety analyses are conducted for Super Fast Reactors (Super FRs), new design concepts of the fourth generation reactors. This study covers the time dependent start-up analysis of a two-flow pass core Super FR and the safety analysis of a single-flow pass core Super FR.
Chapter 1 introduces the necessary of the study. A Super FR is an innovative nuclear reactor concept which is developed based on the technologies of supercritical fossil-fired power plants (FPPs) and light water reactors (LWRs). Once-through coolant cycle is adopted and the plant is operated at supercritical pressure, 25 MPa, with high efficiency of about 45%. The plant system is simple due to the use of once-through direct cycle. Specific heat of the coolant is high at the pseudo-critical temperature so that the heat can be removed by low coolant flow rate. Over these advantages, the start-up and safety characteristics of the innovative reactor, however, should be clarified.
Start-up system of a Super FR will be different from that of the LWRs due to transition of the coolant from subcritical to supercritical pressure. There will be a large change of the coolant properties during start-up. A potential of high maximum cladding surface temperature (MCST) at the critical pressure is to be a concern during the start-up. Procedures of the start-up system should be investigated. Thermal and sensitivity analyses of the start-up system should be carried out to clarify the MCST during start-up.
A single-flow pass core Super FR has been designed to simplify refueling and core structures of the two-flow pass core Super FR. Safety analysis of the core design is important due to its high power density with low coolant flow rate. The safety principle is different from that of LWRs which keeping the coolant flow rate is the fundamental safety requirement instead of keeping the water level. Furthermore, the single-flow pass core has less water inventory inside the reactor pressure vessel (RPV) and introduces an axial large change of coolant density which leads to a high power peaking at the beginning of cycle (BOC).
The flow distribution between seed and blanket assemblies is largely different. It needs a high orifice coefficient in the blanket assembly which might lead to a high pressure drop at flow abnormalities.
Accordingly, even though a safety analysis of the two-flow pass core was carried out, safety performance of the single-flow pass Super FR will be different. Safety analysis of the Super FRshould be clarified.
Meanwhile, passive safety systems of the Super FR are important in case of station blackout. It should be proposed because emergency power is not necessary. Applicability of the passive systems to the Super FR should be investigated.
Objectives of this study are (1) to perform start-up thermal analysis of a two-flow pass core Super FR, (2) to perform safety analysis of a single-flow pass core Super FR, (3) to propose and to analyze a passive safety system of a Super FR.
Chapter 2describes the analyses method. SPRAT codes for start-up and safety analyses are modified based on the core designs of Super FRs.
Chapter 3 explains the results of the start-up analysis. A recirculation system is provided in the start-up system of the two-flow pass core Super FR for pressurization and for raising the coolant temperature. The recirculation system which is separated from the once-through direct cycle consists of a steam drum, a heat exchanger, a recirculation pump, recirculation pipe and valves. Turbines are started with supercritical steam which is physically similar to superheated steam.
After the deaeration, the start-up procedures are divided into 4 phases: (1) raising the feedwater temperature to operating temperature of 280oC, (2) pressurization to 25 MPa, (3) line-switching from the recirculation system to once-through coolant cycle and (4) power-raising to 100%. During subcritical pressure, heat flux of each fuel assembly is low compared to the critical heat flux (CHF) which was calculated based on the 1995 Groeneveld look-up table. The minimum departure from nucleate boiling ratio (MDNBR) is high due to high flow to power ratio. The cladding surface temperature is not sensitive to the change of inlet temperature, recirculation flow rate and gap volume due to the large safety margin. It is the advantage of the start-up system with a recirculation system.
The power is raised to 100% after the line-switching at supercritical pressure. During this phase, safety margin of the first pass (blanket and seed 1) is large due to low coolant temperature, while the safety margin of the second pass (seed 2) is more limited due to high coolant temperature. The minimum flow to power ratio during this phase is 1.0 to keep the MCST of seed 2 low.
Chapter 4discusses the results of the safety analysis of the single-flow pass Super FR. Accident of total loss of feedwater flow is identified as the most important event due to a lack of heat transfer by convection. There is no coolant expansion effect which appears in a two-pass core Super FR. The generated heat is accumulated inside the core leading to a high MCST before coolant injection from auxiliary feedwater system (AFS). However, the MCST is effective to decrease after the coolant injection due to no effect of turbine control valve operation which leads to a high MCST in a two-flow pass Super FR.
Mitigation of the MCST before coolant injection is by actuating the automatic depressurization system (ADS) which actually the signal of flow low 6% is provided for the actuation itself. The MCST is mitigated due to large coolant flow rate induced during the depressurization.
Less water inventory in the RPV leads the pressure to be sensitive to depressurization events. In case of the accident of total loss of feedwater flow without control system actuation, this characteristic leads to an early ADS actuation which mitigates the accident effectively. In case of LOCA accidents, the sensitivity also leads the ADS to be actuated early by pressure low, 23.5 MPa. The MCST is to be much smaller compared to these accidents in a two-flow pass Super FR. However, if the control system is considered, a critical break is introduced at 7.8% break of small LOCA leading to a high MCST. Early ADS actuation, however, is effective to mitigate this accident.
In case of 100% break large LOCA, if one low pressure coolant injection (LPCI) is used for reflooding, a high MCST is introduced due to the high power peaking. It is mitigated effectively by using two LPCIs which inject about 50% of rated flow. The location of the power peaking at the bottom part itself is to be the advantage which fast cooling is necessary only at the beginning. Afterwards, the MCST is decreased rapidly. In view of the core design, the power peaking can be managed by applying different mixed oxide (MOX) fuel enrichment in axial direction. For smaller power peaking, the MCST at large LOCA is less severe.
In case of abnormal transient of loss of offsite power, high orifice coefficient in the blanket assembly causes high pressure drop during coolant injection leading to a late flow generation in the assembly. As a result, the MCST continues to increase. However, low thermal power in the assembly limits the MCST to be lower than the criterion. In case of the event without scram actuation (ATWS), the MCST is increased rapidly due to the high power density of the reactor. The ADS needs to be operated to mitigate the accident.
Characteristic of negative void reactivity can suppress the power during depressurization and the MCST is mitigated.
Chapter 5discusses the passive safety system of a Super FR. Passive safety systems which consist of isolation condenser (ICS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS) are proposed for a Super FR. Safety analysis of the single-flow pass Super FR by using the passive systems shows that the passive systems are applicable to a Super FR.
Accident of total loss feedwater flow rate and large LOCA are mitigated by the passive systems.
Chapter 6summarizes the conclusions. (1) The start-up system of a two-flow pass core Super FR and its procedures are clarified. Large safety margin is the advantage of the start-up system with a recirculation system. (2) Safety characteristics of the single-flow pass Super FR are clarified. Sensitivity to depressurization events is to be the advantage of the reactor because the events will be mitigated by an early ADS operation actuated by pressure low of 23.5 MPa. However, the ADS also needs to be operated in case of ATWS to mitigate the rapid increase of the MCST due to the high power density. (3) Passive system of a Super FR is developed and it is applicable to the Super FR for accident mitigation.
Acknowledgments
The author would like to express his sincere gratitude to Professor Yoshiaki Oka for his guidance, encouragement, and support during the study. Special thanks are for his valuable advice, and for reviewing this thesis. It is truly a privilege to study with him, and benefit from his invaluable expertise.
The author would also like to thank Professor Masakazu Washio for his guidance in preparing this thesis, and also for his valuable advice and comments to this thesis. Thanks are also due to Professor Shinichi Morooka and Professor Seiichi Yokobori for their valuable advices and comments in improvement of writing this thesis.
The author would also like to thank Dr. Triwulan Tjiptono, Dr. Anwar Budianto, Prof. Adhi Susanto, Prof. Kris Tri Basuki and Dr. Supriyono for their support of doing this study. Thanks are also due to Dr.
Haipeng Li and Dr. Qingjie Liu for their help in understanding the codes used for the study, for providing the data of the core design and also for the discussion during the study. Thank is also due to Mrs. Yuki Kikuchi for her help during the study.
Special thanks to my wife, drg. Maria Triananingsih, who has continually provided the inspiration and support necessary for finishing this study and writing this thesis.
The author acknowledges the contributions of the members of Oka Laboratory for useful discussion and information during the study:
- Dr. Asril Pramutadi Andi Mustari - Dr. Ronghua Chen
- Dr. Lite Zhang - Mr. Jianhui Wu - Mr. Gen Li - Mr. Rui Guo - Ms. Xin Li - Mr. Nobuo Tamiya - Mr. Matsuura Takashi - Mr. Takayoshi Kamata - Mr. Motohiro Sakakibara - Mr. Takeo Watanabe
Financial support from the Ministry of Research and Technology of Indonesia is gratefully acknowledged.
4 6>: EBHA7 8I8?BC@ 8AGB9,HC8E + BE8 8F<: A
-JB9?BJ C4 FF6BE8,HC8E +
,<A: ?89?BJ C4 FF6BE8,HC8E +
) I8EI<8J B9* 4 FG,GH7<8F
,G4 EGHCFLFG8@ FB9,HC8E +
,4 98GL4 A4 ?LF<FB9,HC8E +
* 4 FF<I8F4 98GLFLFG8@ B9,HC8E + ' BG<I4 G<BAF4 A7 ) 5=86G<I8FB9,GH7L - ;8F<F) E: 4 A<M4 G<BA
,4 98GL A4 ?LF<F B78F 4 ?6H?4 G<BA ' B78?
4 ?6H?4 G<BA @ B78? B9FG4 EGHC4 A4 ?LF<F 4 ?6H?4 G<BA @ B78? B9F4 98GL4 A4 ?LF<F ' 4 FF4 A7 A8E: L BAF8EI4 G<BA ' B78?F + 4 7<4 ? " 84 G-E4 AF98E' B78?
* B<AG% <A8G<6F' B78?
' B@ 8AGH@ BAF8EI4 G<BA ' B78?
" 84 G-E4 AF98E BE@ H?4 " 84 GGE4 AF98E4 GFHC8E6E<G<64 ? CE8FFHE8 " 84 GGE4 AF98E4 GFH56E<G<64 ? CE8FFHE8 8G8E@ <A4 G<BA B96E<G<64 ? 984 G9?HK "
E<G<64 ? ?BJ 4 ?6H?4 G<BA4 A7 + 89?BB7<A:
* ?4 AG,G4 EG. C,LFG8@
,G84 @ EH@ 8F<: A ,G4 EGHC* EB687HE8F - ;8E@ 4 ? A4 ?LF<F B9- ;8 ,G4 EG. C ,LFG8@ " 84 G 9?HK @ 4 E: <A 7HE<A: FH56E<G<64 ? CE8FFHE8
,G4 EGHCF8AF<G<I<GL
- ;8E@ 4 ? 4 A4 ?LF<F7HE<A: CBJ8EE4 <F<A:
,H@ @ 4 EL
* ?4 AG LA4 @ <64 A7 BAGEB? ,LFG8@ 8F<: A B9- ;8,HC8E + * ?4 AG7LA4 @ <6J<G;BHG6BAGEB? FLFG8@
BAGEB? FLFG8@ 78F<: A
,4 98GL,LFG8@ 8F<: A ,4 98GL E<G8E<4 I8AGF,8?86G<BA 66<78AGF 66<78AGFJ<G; 6BAGEB? FLFG8@
66<78AGFJ<G;BHG6BAGEB? FLFG8@
5ABE@ 4 ? -E4 AF<8AGF
& BFFB9 BB?4 AG 66<78AG& )
,@ 4 ?? & )
& 4 E: 8& )
AG<6<C4 G87 -E4 AF<8AG0 <G;BHG,6E4 @ - 0 , ,H@ @ 4 EL
* 4 FF<I84 A4 ?LF<F6B78
#FB?4 G<BA 6BA78AF8EFLFG8@ 4 A7 6BE8@ 4 >8HCG4 A>
' B78? 78I8?BC@ 8AG
4 ?6H?4 G<BA E8FH?GFB9<FB?4 G<BA 6BA78AF8EFLFG8@ 4 A7 6BE8@ 4 >8HCG4 A>
! E4 I<GL E<I8A BB?<A: ,LFG8@
* 4 FF<I8 BAG4 <A@ 8AG BB?<A: ,LFG8@
<: ( H@ 58EB9AH6?84 ECBJ8EC?4 AGF<A BC8E4 G<BA JBE?7J<78 4 FB9$ 4 AH4 EL <: *;4 F87<4 : E4 @ B9J4 G8E <: ;4 A: 8B978AF<GL4 GFH56E<G<64 ? 4 A7 FHC8E6E<G<64 ? CE8FFHE8 <: ;4 A: 8B9FC86<9<6;84 G4 GFH56E<G<64 ? 4 A7 FHC8E6E<G<64 ? CE8FFHE8 <: ) A68G;EBH: ; 7<E86G6L6?8B94 ,HC8E + <: ,6;8@ 4 G<6B9,HC8E + <: ?BJ F6;8@ 4 G<6B9G;8GJBC4 FF,HC8E + J<G; 4 ?? HCJ4 E7 9?BJ <: & 4 LBHG4 FF8@ 5?<8FB94 F<A: ?8C4 FF,HC8E + <: * EB9<?8FB9CBJ8E7<FGE<5HG<BA B9G;8F<A: ?8C4 FF,HC8E + <: ;4 E4 6G8E<FG<6FB9F?<7<A: CE8FFHE8FG4 EGHCB94 , 0 + <: ' 4 K<@ H@ 6?4 77<A: FHE94 68G8@ C8E4 GHE87HE<A: FG4 EGHCB94 , 0 + <: ( 8J 78F<: A B9FG4 EGHCF6;8@ 8 <: ,4 98GLFLFG8@ B9,HC8E& 0 + <: -BG4 ? ?BFFB99?BJ <A G;8GJB9?BJ C4 FF,HC8E + J<G; 4 ?? HCJ4 E7 9?BJ <: BB?<A: FLFG8@ B9C?4 AGFLFG8@ F <: 4 F<6FGEH6GHE8B9,* + - 6B78 <: 4 ?6H?4 G<BA @ B78? 9BEFG4 EGHCFLFG8@ 4 A4 ?LF<FB9G;8GJBC4 FF,HC8E + <: ?BJ6;4 EGB9FG4 EGHCG;8E@ 4 ? 4 A4 ?LF<F9BEG;8,HC8E + <: 4 ?6H?4 G<BA @ B78? B9F4 98GL4 A4 ?LF<FB9G;8F<A: ?89?BJ C4 FF6BE8,HC8E + <: ?BJ6;4 EGB9F4 98GL4 A4 ?LF<F6B789BEG;8F<A: ?8C4 FF,HC8E + <: 4 ?6H?4 G<BA @ B78? 9BEE89?BB7<A: C;4 F8B9G;8F<A: ?89?BJ C4 FF6BE8,HC8E + <: ?BJ6;4 EGB9E89?BB7<A: 64 ?6H?4 G<BA <: ;4 A: 8B9FG84 @ 9?BJ E4 G84 F9HA6G<BA B9GHE5<A86BAGEB? I4 ?I8BC8A<A:
<: 887J4 G8E9?BJ E4 G84 F9HA6G<BA B96BE8CE8FFHE8 <: ' B78? B9E4 7<4 ? ;84 GGE4 AF98E9EB@ 9H8? C8??8GGB 6BB?4 AG <: + 8: <BAFB9;84 GGE4 AF98E4 GG;89H8? 6;4 AA8? 4 GFH56E<G<64 ? CE8FFHE8 <: + 8: <@ 8FB9CBB? 5B<?<A:
<: BHA74 E<8F4 A7 9?BJ 7<E86G<BA <: ?BJ C4 GG8EA B9G;8GJBC4 FF,HC8E + J<G; 4 ?? HCJ4 E7 9?BJ <: ,G4 EGHCFLFG8@ B9G;8,HC8E 4 FG+ 84 6GBE <: ,G4 EGHC6HEI89BE,HC8E + <: + 4 <F<A: B99887J4 G8EG8@ C8E4 GHE8GB ^ 5L6BAFG4 AGF@ 4 ?? CBJ8E <: ,6;8@ 8B9CE8FFHE<M4 G<BA 7HE<A: FG4 EGHC <: HEI8B96BAGEB??<A: J4 G8E?8I8? 7HE<A: CE8FFHE<M4 G<BA <: 0 4 E@ <A: B9G;8FG84 @ ?<A8
<: ) C8A<A: GHE5<A85LC4 FFI4 ?I8F <: ) C8A<A: B9@ 4 <AFG84 @ FGBC4 A7 GHE5<A86BAGEB? I4 ?I8F <: ,;HG7BJA B9G;86<E6H?4 G<BAFLFG8@
<: + <F8B9CE8FFHE84 F9HA6G<BA B9G;8CBJ8E <: " 84 G9?HK@ 4 E: <A 7HE<A: FH56E<G<64 ? CE8FFHE8 <: ( + B99H8? 6;4 AA8?F7HE<A: CE8FFHE<M4 G<BA <: 4 ( + B95?4 A>8G4 FF8@ 5?L4 G ' *4 <: ' ,- 7HE<A: FH56E<G<64 ? CE8FFHE8 <: #A9?H8A68B9<A?8GG8@ C8E4 GHE87HE<A: CE8FFHE<M4 G<BA <: #A9?H8A68B9: 4 CIB?H@ 87HE<A: CE8FFHE<M4 G<BA <: #A9?H8A68B99?BJ E4 G87HE<A: CE8FFHE<M4 G<BA <: ' ,- B9G;89<EFGF887 4 FF8@ 5?L7HE<A: CBJ8EE4 <F<A: C;4 F8 <: ' ,- B9G;8F86BA7 F887 4 FF8@ 5?L7HE<A: CBJ8EE4 <F<A: C;4 F8 <: ' ,- B95?4 A>8G4 FF8@ 5?L7HE<A: CBJ8EE4 <F<A: C;4 F8 <: BB?4 AG78AF<GL6;4 A: 8<A G;8F<A: ?8C4 FF,HC8E + <: B@ C4 E<FBA B96BB?4 AG78AF<GL6;4 A: 858GJ88A G;8F<A: ?8C4 FF4 A7 GJBC4 FF6BE8F <: I4 E4 : 8;BG6;4 AA8? 4 GABE@ 4 ? BC8E4 G<BAB9 ) <: + 8FCBAF8B9G;8,HC8E + GB FG8CJ<F8E84 6G<I<GL<AF8EG<BA <: + 8FCBAF8B9G;8,HC8E + GB FG8CJ<F8786E84 F8<A 9887J4 G8E9?BJ E4 G8 <: + 8FCBAF8B9G;8,HC8E + GB FG8CJ<F8786E84 F8<A GHE5<A86BAGEB? I4 ?I8BC8A<A:
<: ,6;8@ 8B96BAGEB? FLFG8@ 9BEG;8,HC8E + <: * E8FFHE86BAGEB? FLFG8@ B9G;8,HC8E + <: 4 ?6H?4 G<BA E8FH?GF9BEGHA<A: : 4 <A <AG;8CE8FFHE86BAGEB? FLFG8@ B9G;8,HC8E + <: ,6;8@ 8B9@ 4 <AFG84 @ G8@ C8E4 GHE86BAGEB? FLFG8@
<: 4 ?6H?4 G<BA E8FH?GF9BEGHA<A: CEBCBEG<BA4 ? : 4 <A <A @ 4 <A FG84 @ 6BAGEB? FLFG8@ B9 G;8,HC8E + <: 4 ?6H?4 G<BA E8FH?GF9BEGHA<A: <AG8: E4 ? : 4 <A <A @ 4 <A FG84 @ G8@ C8E4 GHE86BAGEB?
FLFG8@
<: ,6;8@ 8B9CBJ8E6BAGEB? FLFG8@ 9BEG;8,HC8E + <: 4 ?6H?4 G<BAE8FH?GF9BEGHA<A: @ 4 K<@ H@ 78I<4 G<BA9BECEBCBEG<BA4 ? 6BAGEB? <AE84 6GBE
CBJ8E6BAGEB? FLFG8@ B9G;8,HC8E + <: ,4 98GLFLFG8@ B9G;8F<A: ?89?BJ C4 FF6BE8,HC8E + <: ,6E4 @ F<: A4 ?F <: -BG4 ? ?BFFB99?BJ J<G;BHG , 4 6GH4 G<BA <: B@ C4 E<FBA B9GBG4 ? ?BFFB99887J4 G8E9?BJ 58GJ88A G;8F<A: ?84 A7 GJBC4 FF8F
,HC8E + <: ,8AF<G<I<GL4 A4 ?LF8F9BEGBG4 ? ?BFFB99887 J4 G8E9?BJ
<: -BG4 ? ?BFFB99?BJ J<G; , 4 6GH4 G<BA <: + 84 6GBE6BB?4 AGCH@ CF8<MHE8 <: BAGEB? EB7 8=86G<BA ;BGFG4 A75L <: BAGEB? 8=86G<BABC8E4 G<BA <: -BG4 ? ?BFFB99887J4 G8E9?BJ J<G;BHG6BAGEB? FLFG8@
<: + 84 6GBE6BB?4 AGCH@ CF8<MHE8J<G;BHG6BAGEB? FLFG8@
<: BAGEB? EB7 8=86G<BA J<G;BHG6BAGEB? FLFG8@ ;BGFG4 A75L <: BAGEB? EB7 8=86G<BA J<G;BHG6BAGEB? FLFG8@ BC8E4 G<BA <: & BFFB99887J4 G8E;84 G<A:
<: #A4 7I8EG8AGFG4 EGHCB9 , <: * 4 EG<4 ? ?BFFB99887 J4 G8E9?BJ <: & BFFB9B99F<G8CBJ8E <: & BFFB9GHE5<A8?B4 7 J<G; 5LC4 FFBC8A <: & BFFB9GHE5<A8?B4 7 J<G;BHG5LC4 FFI4 ?I8FBC8A <: . A6BAGEB??87 J<G;7E4 J4 ? B94 6BAGEB? EB7 ;BGFG4 A75L <: BAGEB? EB7 4 5ABE@ 4 ?<GLCH??BHGBC8E4 G<BA <: 887 J4 G8E9?BJ E4 G86BAGEB? FLFG8@ 94 <?HE8 <: * E8FFHE86BAGEB? FLFG8@ 94 <?HE8 <: ' ,#/ 6?BFHE8 <: ,@ 4 ?? & ) J<G;BHG6BAGEB? FLFG8@
<: ' ,- B9F@ 4 ?? & ) J<G;BHG6BAGEB? FLFG8@
<: * E8FFHE84 G6B?7?8: F@ 4 ?? 5E84 >& ) J<G; 6BAGEB? FLFG8@
<: ,@ 4 ?? & ) 4 GHCC8E?<@ <GB9F<A: ?89?BJ C4 FF6BE84 A7 4 GJBC4 FF6BE8,HC8E + F <: ' ,- B9F@ 4 ?? & ) J<G; 6BAGEB? FLFG8@
<: ' ,- 4 G6E<G<64 ? 5E84 >B9F@ 4 ?? & ) ) 5E84 > 4 A7 ) 5E84 >
<: 9986GB984 E?L , 4 6GH4 G<BA G<@ 8BA G;86E<G<64 ? ' ,- <A F@ 4 ?? & ) <: 4 E?L , 4 6GH4 G<BA B9F@ 4 ?? & ) 4 G6E<G<64 ? 5E84 >F <: * E8FFHE84 G6B?7?8: ?4 E: 8& ) <: H8? 6;4 AA8? 9?BJ E4 G84 A7 @ 4 K<@ H@ 6?4 77<A: G8@ C8E4 GHE84 G6B?7?8: ?4 E: 8& )
<: & 4 E: 8& ) B9 5E84 >F <: + 89?BB7<A: B9 5E84 >5LHF<A: & * # <: + 89?BB7<A: B9 5E84 >& ) 4 G ) <: + 89?BB7<A: C;4 F8B96B?7?8: 5E84 >& ) <: & BFFB99887 J4 G8E;84 G<A:
<: + 84 6G<I<GL988754 6>B9?BFFB99887 J4 G8E;84 G<A:
<: - 0 , B9C4 EG<4 ? ?BFFB99887 J4 G8E9?BJ <: + 84 6G<I<GL988754 6><A C4 EG<4 ? ?BFFB99887 J4 G8E9?BJ <: - 0 , B9?BFFB9B99F<G8CBJ8E <: + 84 6G<I<GL988754 6>B9?BFFB9B99F<G8CBJ8E <: - 0 , B9?BFFB9GHE5<A8?B4 7 J<G;BHG5LC4 FFBC8A <: - 0 , B96BAGEB? EB7 4 5ABE@ 4 ?<GLCH??BHG;BGFG4 A75L <: + 84 6G<I<GL988754 6>B96BAGEB? EB7 4 5ABE@ 4 ?<GLCH??BHG;BGFG4 A75L <: BAGEB? EB7 4 5ABE@ 4 ?<GLCH??BHGBC8E4 G<BA <: + 84 6G<I<GLB96BAGEB? EB7 4 5ABE@ 4 ?<GLCH??BHGBC8E4 G<BA <: - 0 , B9CE8FFHE86BAGEB? FLFG8@ 94 <?HE8 <: - 0 , B9' ,#/ 6?BFHE8 <: ,H@ @ 4 ELB9' ,- 9BE4 ?? ABA& ) 8I8AGF <: ,H@ @ 4 ELB9C84 >CE8FFHE89BE4 ?? ABA& ) 8I8AGF <: ,H@ @ 4 ELB9' ,- 9BE& ) 4 66<78AG <: ,H@ @ 4 ELB9' ,- 9BE -0 , 4 66<78AGF <: ,H@ @ 4 ELB9CE8FFHE89BE - 0 , 4 66<78AGF <: *4 FF<I8F4 98GLFLFG8@ B94 ,HC8E + <: BA9<: HE4 G<BA B9<FB?4 G<BA 6BA78AF8EFLFG8@
<: 4 ?6H?4 G<BA @ B78? B9G;8 ' - <: 4 ?6H?4 G<BA @ B78? 9BE# , B9G;8F<A: ?89?BJ C4 FF6BE8,HC8E + <: ?BJ6;4 EGB99?BJ 64 ?6H?4 G<BA 4 GFHC8E6E<G<64 ? CE8FFHE8 <: 4 ?6H?4 G<BA @ B78? B96BE84 A7 6BA78AF8ECE8FFHE8 <: ?BJ6;4 EGB96BE8CE8FFHE864 ?6H?4 G<BA <: 4 ?6H?4 G<BA @ B78? B9A4 GHE4 ? 9?BJ 64 ?6H?4 G<BA <: E<I<A: 9BE684 A7 CE8FFHE87EBC4 G9<K87 9?BJ E4 G8 <: * E8FFHE87EBCB984 6; ?<A84 ?BA: G;8<FB?4 G<BA 6BA78AF8E <: E<I<A: 9BE684 A7 CE8FFHE87EBC4 G9<K87 <A?8GG8@ C8E4 GHE8 <: -BG4 ? ?BFFB99887J4 G8E9?BJ 9BE7<998E8AGGH58FAH@ 58E <: -BG4 ? ?BFFB99887J4 G8E9?BJ 9BE7<998E8AG;8<: ;GB9G;8# , <: ;4 E4 6G8E<FG<6FB9G;8 , 4 6GH4 G<BA B9G;8F<A: ?89?BJ C4 FF,HC8E + <: & <A8FJ<G6;<A: 9EB@ , GB# , ?<A8F <: 4 ?6H?4 G<BA E8FH?GFB9CE8FFHE8 <: 4 ?6H?4 G<BA E8FH?GFB9FG84 @ 9?BJ <: 4 ?6H?4 G<BA E8FH?GFB9G;86BA78AF4 G89?BJ <: & BA: G8E@ 6BB?<A: 5L# , <: ;4 E4 6G8E<FG<6FB9 ' - <: ' <G<: 4 G<BA B9' ,- 5L ' -
<: ,6;8@ 4 G<6B9! , <: + 89?BB7<A: 5L! , <: 0 4 G8E?8I8? B9<A78C8A78AG! , 9BEE89?BB7<A:
<: BB?4 AG<A=86G<BA FC887 B9<A78C8A78AG! , 9BEE89?BB7<A:
<: 0 4 G8E?8I8? 4 A7 BHG?8G9?BJ E4 G8B9! , 7HE<A: E89?BB7<A:
<: + 89?BB7<A: 5L! , <: 4 ?6H?4 G<BA @ B78? B9C4 FF<I86BAG4 <A@ 8AG6BB?<A: FLFG8@
List of Table
-4 5?8 ;4 E4 6G8E<FG<6FB9G;8GJBC4 FF,HC8E + J<G; 4 ?? HCJ4 E7 9?BJ -4 5?8 BE86;4 E4 6G8E<FG<6FB9BA8C4 FF,HC8E + -4 5?8 B@ C4 E<FBA B9GE4 AF<8AGF4 A7 4 66<78AGF -4 5?8 * 4 E4 @ 8G8EFB9E84 6GBE7LA4 @ <6FB9,HC8E + -4 5?8 " 84 GGE4 AF98E6BEE8?4 G<BAFHF87 9BEFH56E<G<64 ? CE8FFHE8 -4 5?8 ,8?86G<BA B99?BJ ;84 GGE4 AF98E6BEE8?4 G<BAF4 GFH56E<G<64 ? CE8FFHE8 -4 5?8 ,8?86G<BA B9CBB? 5B<?<A: ;84 GGE4 AF98E6BEE8?4 G<BAF -4 5?8 ;4 E4 6G8E<FG<6FB9G;8J4 G8EF8C4 E4 GBEB9E898E8A68FHC8E6E<G<64 ? 5B<?8E -4 5?8 B@ C4 E<FBA B9FG84 @ 7EH@ 7<@ 8AF<BAHF87 <A FG4 EGHCB9,HC8E + -4 5?8 CC?<64 5?8E4 A: 8FB9G;8! EB8A8I8?7 ?BB>HCG4 5?8 -4 5?8 ,8AF<G<I<GL4 A4 ?LF<F7HE<A: CE8FFHE<M4 G<BA -4 5?8 ,4 98GLFLFG8@ 4 6GH4 G<BA -4 5?8 66<78AGF4 A7 GE4 AF<8AGF9BEF4 98GL4 A4 ?LF<F -4 5?8 6GH4 G<A: F<: A4 ? 4 A7 7E<I<A: 9BE68B9G;8C4 FF<I8FLFG8@
-4 5?8 4 C4 6<GLB96BB?4 AG<A + */ B9G;8F<A: ?89?BJ C4 FF,HC8E +
1. Introduction
1.1 Background
Primary energy consumption is increasing every year in line with the increase of LKLQH=PEKJEJPDA SKNH@ (J PDA SKNH@]OLNEI=NU AJANCU ?KJOQILPEKJS=OHAOOPD=J !PQ and increased to be more than 11 Btu in 2009 (Oka, 2013). Fossils which are oil and coals are still being the dominant energy sources which are giving impacts on environment in the form of global warming due to the release of CO2from the fossil-fired power plants. In the past few decades, efforts were made to improve the thermal efficiency of the power plants in order to resolve the global environment issues and in order to reduce the cost of fuel consumption (Yi, 2004).
Use of nuclear energy does not release CO2 and other toxic gases into environment. It contributed about 5.5% of the total worldys energy consumption in 2009 (Oka, 2013). The number of nuclear power plants in operation as of January 2013 is shown in Fig. 1-1. Thermal efficiency of the nuclear power plants, however, is important to be increased to improve its competitive against the fossil fuel power plants.
Nuclear power plants are developed in line with the development of boilers which have evolved from primitive boilers to circular boilers and once-through boilers (Oka et al., 2010).
The current light water reactors (LWRs) which are boiling water reactors (BWRs) and pressurized water reactors (PWRs) adopt the circular boilers operated at subcritical pressure.
Although the nuclear reactors are already well proven in industries, however, improvement of the efficiency is difficult to be achieved significantly, especially due to limitation of boiling point at subcritical pressure. Meanwhile, the type of once-through boilers has been adopted in the current fossil fuel power plants which are operated at supercritical pressure. Advanced steam conditions of 25 MPa/600 oC/610 oC have been possible in the power plants and the plant efficiency can be increased significantly up to 45% (Yi, 2004a). Therefore, one way to increase the efficiency of nuclear power plants is to use the advanced supercritical steam condition. The supercritical steam can be generated in a nuclear power plant by adopting the once-through direct cycle operated at supercritical pressure.
Fig. 1-1 Number of nuclear power plants in operation, worldwide, as of January 2013 (www.euronuclear.org, accessed on February 2014)
Use of supercritical steam cycle in nuclear power plants will give some advantages over the current existing LWRs due to some excellent water properties at supercritical pressure.
Characteristics of water at supercritical pressure compared to that at subcritical pressure are shown in Figs. 1-2, 1-3, and 1-4 (Oka et al., 2010; Oka, 2013). The critical pressure of water is at 22.1 MPa. At supercritical pressure, the coolant does not experience a change of phase.
The coolant density decreases continuously with the coolant temperature. It is different from that at subcritical pressure which the density changes abruptly. There is no boiling during operation so that the outlet coolant temperature can be raised without the limit of boiling point.
It means that the efficiency can be increased significantly.
Fig. 1-2 Phase diagram of water (Oka et al., 2010)
Fig. 1-3 Change of density at subcritical and supercritical pressure (Oka et al., 2010)
Fig. 1-4 Change of specific heat at subcritical and supercritical pressure (Oka et al., 2010)
The specific heat is high at the pseudo-critical temperature which the heat can be removed efficiently. The coolant enthalpy difference before and after the pseudo-critical temperature is large so that the heat can be removed by low coolant flow rate. It is about one-fifth to one-tenth of LWRs. Adoption of once-through direct cycle cooling system leads the plant system to be simple due to absence of recirculation system and dryers of BWRs.
Steam generators of PWRs are also not used.
The other advantage is the good compatibility of the once-through plant with a tight fuel lattice fast reactor core. The plant system configuration can be identical for both fast and thermal reactors (Oka, 2010).
In Japan, research and developments (R&Ds) of supercritical light water reactors have been carried out since 1989 at the University of Tokyo and Waseda University (Oka et al., 2010). The new reactor concepts are called as Super Light Water Reactors (Super LWRs) and Super Fast Reactors (Super FRs). The Super FR is the fast version of the Super LWR.
In the world, the new concept of nuclear reactors is called commonly as the supercritical water cooled reactors (SCWRs). R&Ds of the SCWRs are also conducted in some other countries throughout the world like Germany, USA, Canada, Russia, Korea and China.
Collaboration of the R&Ds is well known as the Generation Four International Forum (GIF) which was started in 2002.
1.2 Development of Super FR Core Design
Two phases of R&Ds of the Super FRs have been conducted at the University of Tokyo and Waseda University (Oka et al., 2011; Oka et al., 2013). A Super FR was designed to find a way of competitive plutonium utilization. It has some advantages over the thermal version one, Super LWR. A Super FR has higher power density which will improve the thermal efficiency and the capital cost. There is also potential of fuel breeding. The plant system of a Super FR is shown in Fig. 1-5.
The design challenge of a Super FR is the possibility of void reactivity coefficient to be positive during loss of coolant. Therefore, a Super FR must be designed by ensuring the total reactivity coefficient to be negative in case of accidents (Li and Oka, 2013). Design of a blanket assembly using zirconium hydride (ZrH) layer was introduced to achieve negative void reactivity (Oka et al., 1996; Cao et al., 2008). Furthermore, flow pattern of the core has been improved for simplification the core design (Liu and Oka, 2013a).
Three designs of Super FRs have been developed as shown in Fig. 1-6 (Yoo et al., 2006;
Oka et al., 2013; Liu and Oka, 2013a; Liu and Oka, 2013b). Fig. 1-6(1) shows a two-flow pass core design of the Super FR with downward-upward flow. The downward coolant
passing from the upper dome of the reactor pressure-vessel (RPV) made the upper core structure of the two-flow pass core to be complex. For simplification of the structure of upper core, a two-flow pass core design of the Super FR with all-upward flow in all assemblies was designed as shown in Fig. 1-6(b). Then, for simplification of refueling and the structures of lower and gap plenums, a single-flow pass core was designed as shown in Fig. 1-6(c).
Fig. 1-5 Once-through direct cycle of a Super FR (Oka et al., 2010)
Fig. 1-6 Schematic of Super FR
(a) two-flow pass core with downward-upward flow,(b) two-flow pass core with all upward flow, (c) all upward single flow pass core
(Yoo et al., 2006; Oka et al., 2013; Liu and Oka, 2013)
1.2.1 Two-flow pass core Super FR
The two-flow pass Super FR with all-upward flow scheme was designed based on the previous core design of two-pass core with downward-upward flow. Similar fuel and assembly configurations were applied. The ZrH layer was applied in the blanket assembly to improve the void reactivity performance. The hexagonal seed and blanket assemblies were then arranged in the core region to form the required flow scheme. The schematic of flow pattern is shown in Fig. 1-7 (Li and Oka, 2013; Liu and Oka, 2013a).
Fig. 1-7 Flow schematic of the two-pass Super FR with all upward flow (Li and Oka, 2013)
The water flows from feedwater line to the bottom dome through the down comer. Then the coolant flows upward through the first pass to cool the blanket assemblies and some of the seed assemblies (Seed 1). After the first pass, the coolant mixes in the gap plenum above the Blanket and Seed 1, and then goes downward through the gap space between assembly boxes and core shroud to the lower plenum. The coolant then flows upward to cool the rest of the seed assemblies (Seed 2) and finally is delivered to the main steam line through the upper plenum. The characteristics of the core design are summarized in Table 1-1.
Table 1-1 Characteristics of the two-pass Super FR with all upward flow (Liu and Oka, 2013a)
Parameters Value
Thermal/electrical power (MWt/MWe) 2325/1000
Pressure (MPa) 25
Inlet & outlet temperature (oC) 280/500
Active core height (m) 1.8
Power density (MW/m3) 171
Number of assemblies for Seed1/Seed2/Blanket 72/126/97 Number of fuel rods per assembly for Seed 1/Seed 2/Blanket 378/378/61
Fuel rod diameter (mm) 5.5
Total flow rate (kg/s) 1200
Flow distribution between Seed1 and Blanket (%) 80/20
Working pressure (MPa) 25.0
MCST (oC) 640
1.2.2 Single-flow pass core Super FR
The two-flow pass Super FR has some drawbacks which the control rods only put in the second pass due to the structure of upper mixing plenum above the first pass. Structure of the lower mixing is also complicated due to seals between hot and cold coolant in the upstream of fuel assemblies. To simplify refueling and the structures of upper and lower mixing plenum, the single-flow pass core Super FR has been designed. The core characteristics are shown in Table 1-2.
The structures of upper and lower mixing plenums, procedures of refueling, and coolant flow scheme of the single-pass Super FR are simple as those of PWR (Liu and Oka, 2013b).
The coolant with temperature of 280 oC flows into the lower plenum through a downcomer.
Then the coolant is distributed into two fuel assemblies which 92.3% of the coolant flows into seed assembly and 7.7% flows into blanket assembly. Afterwards the coolant mixes in the upper plenum and flows out through hot-leg with temperature of 500oC. The flow distribution is performed by orifices. The large difference of flow distribution, however, needs a large orifice coefficient in the blanket assembly. In case of flow rate abnormality, high pressure drop might appear in the blanket assembly. It might also influence the cooling of the assembly during the abnormality
Table 1-2 Core characteristics of one-pass Super FR (Liu and Oka, 2013b)
Parameters Value
Thermal/electrical power (MWt/MWe) 2353/1000
Pressure (MPa) 25
Inlet/outlet temperature (oC) 280/500
Active core height (m) 2.4
Equivalent core diameter (m) 2.45
Number of assembly (Seed/Blanket) 78/37 Number of fuel rods/assembly (Seed/Blanket) 978/547
Total flow rate (kg/s) 1203
Flow distribution of Seed/Blanket (%) 92.3/7.7 MCST of Seed channel BOC/EOC (oC) 646/647 MCST of Blanket channel BOC/EOC (oC) 617/647 .
Fig. 1-8 Layout assemblies of a single-pass Super FR
(a) seed assembly, (b) and (c) upper and lower parts of blanket assembly (Liu and Oka, 2013b)
The seed assembly is comprised of mixed oxide (MOX) fuel as shown in Fig. 1-8(a). It has the function as the primary power generating source. An axial large density change is introduced in the assembly due to the pattern of single-flow pass core. The large density change leads to a high power peaking at the bottom part of the core at the beginning of cycle (BOC) as shown in Fig. 1-9. The high power peaking will influence the safety performance.
Although the enrichment of the MOX can be adjusted, however, it might influence the safety performance during abnormality. Meanwhile, plutonium (Pu) building-up due to the breeding in the UO2zones of the blanket assembly as shown in Fig. 1-8(b) also introduces great power rise in blanket assembly, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at BOC. To address these issues, some MOX fuels are applied in the bottom region of the blanket assembly as shown in Fig 1-8(c). These are to mitigate the power change of the blanket and to increase the outlet temperature of the blanket during cycle. ZrH rods are applied surrounding the breeding region to provide moderation during coolant void. Some SUS rods are arranged at peripheral region of the blanket assembly to reduce the pin-power peaking induced by ZrH to neighboring seed. Regarding these characteristics of the core design, the safety characteristics of the single-flow pass core Super FR should be clarified. Coolant expansion which mitigates an accident of total loss of flow in the two-flow pass core is not expected in the single-flow pass core due to the flow pattern.
Fig. 1-9 Profiles of power distribution of the single-pass Super FR (Liu and Oka, 2013b)
1.3 Overview of Past Studies 1.3.1 Start-up systems of Super FR
Due to the different operating pressure of a Super FR from that of LWRs, two types of start-up systems of a Super FR were designed by referring to start-up systems of fossil fuel power plants (Nakatsuka et al., 2001). The first type was constant pressure start-up system which nuclear heating was started at supercritical pressure. A bypass system which was comprised of a flash tank and pressure-reducing valves was used to establish sufficient flow rate to prevent overheating of fuel cladding and to ensure adequate core cooling during initial operation. However, system configuration of the start-up system was complicated and pressure ramp-up operation was required. The start-up valves would experience a large pressure difference during bypass operation, causing faster erosion, which in turn requires frequent valve maintenance (Oka et al., 2010).
To meet the problems in the constant pressure start-up system, a type of sliding pressure start-up system was designed. The purpose was also to meet the requirements of frequent load cycling. Thermal analysis of the sliding start-up system applied to a Super LWR was studied (Yi et al., 2004). Nuclear heating started at subcritical pressure. Pressurization was assumed independent from the nuclear heating.
Fig. 1-10 Characteristics of sliding-pressure start-up of a SCWR (Yi et al., 2004)
Turbines were also started at subcritical pressure. A startup bypass system consisting of a steam-water separator, a drain tank and drain valves was necessary for two-phase flow at subcritical pressure. Pressurization was done by a pressurizer pump while the reactor power was kept constant. Characteristics of the start-up scheme during pressurization are shown in Fig. 1-10 (Yi et al., 2004). There was a maximum allowable power to satisfy the criterion of maximum cladding surface temperature (MCST) and to prevent water rods from boiling in the Super LWR. The margin at critical pressure was narrow. Characteristics of the MCST during pressurization are shown in Fig. 1-11. At critical pressure, the MCST was high because of high power to satisfy the required turbine inlet steam enthalpy during subcritical pressure. In other side, low flow rate was necessary to minimize the volume of the steam-water separator, but it should be high enough to prevent water rods from boiling in the Super LWR. The safety margin was narrow by using this start-up system. Besides that, the change of core pressure was given with an assumption that the core pressure could be raised independently from the power. The assumption might be too simplified. Since the turbine will be similar to that of fossil-fuel power plant, it should be warmed and started using superheated steam instead of saturated steam. In the start-up system, there is a concern that saturated steam cannot be used for warming and starting the supercritical turbines while the Super LWR has no superheater.
Fig. 1-11 Maximum cladding surface temperature during start-up of a SCWR (Yi et al., 2004)
To meet the problems in the sliding start-up system and requirement of superheated steam for turbine start-up, a new startup scheme of a Super FR has been designed by referring to startup systems of fossil fuel power plants and boiling water reactors (BWRs) (Yamada et al., 2009). The startup scheme uses a recirculation system separated from the once-through direct cycle the same as that of fossil fuel power plant. The start-up system is shown in Fig.1-12 (Oka et al., 2010). Nuclear heating is used for pressurization the same as that in BWRs. Turbines are started at supercritical pressure. In that study the detail thermal analysis of the start-up system, however, was not carried out yet. Potential of high MCST at the critical pressure should be clarified. The sensitivity analysis during pressurization and the thermal analysis during power-raising were also not studied yet.
Fig. 1-12 New design of start-up scheme (Oka et al., 2010)
1.3.2 Safety analysis of Super FR
Some studies of safety analysis of Super LWRs and Super FRs were carried out in the past time (Ishiwatari et al., 2005; Ikejiri et al., 2010; Li and Oka, 2013, Tamiya et al., 2014;
Momomi, 2014). A safety system of a Super LWR was designed based on the fundamental of safety requirement of once-through coolant cycle which is keeping the coolant flow rate instead of keeping the water level. The designed safety system is shown in Fig. 1-13 (Ishiwatari et al., 2005a).
Abnormality events which are possible to occur in supercritical nuclear reactor were also explained (Ishiwatari et al., 2005b). Abnormality events for the Super LWR were taken
from those of PWR and BWR considering the features of the once-through cooling system.
Comparison of abnormal transients and accidents among PWR, BWR, Super LWR and Super FR is shown in Table 1-3.
Fig. 1-13 Safety system of Super LWR (Ishiwatari et al., 2005a)
Safety analyses of Super FRs were applied for the two-pass core designs (Ikejiri et al., 2010; Li and Oka, 2013). Safety analysis of the two-flow pass core Super FR with downward-upward flow showed that the accident of total loss of feedwater flow was identified as important accident at supercritical pressure. The cladding temperature was the highest among the other events. However, there was reverse flow in the downward flow fuel channels due to low flow rate and buoyancy during the accident. The flow rate of the blanket fuel channel stayed small, the coolant temperature increased and volume expansion occurred.
The coolant expansion leaded to a flow rate into the second pass of the fuel channel. As a result the flow of the second pass was maintained even before coolant injection from AFS.
The MCST of the fuel channel was mitigated. This effect was called aswwater source effect of blanketx(Ikejiri et al., 2010).
Table 1-3 Comparison of transients and accidents
(&: abnormal transients,(: accidents) (Ishiwatari et al., 2005b)
Abnormality
type Event PWR BWR Super LWR and
Super FR Reactivity
abnormality
' Uncontrolled CR withdrawal at start-up
' Uncontrolled CR withdrawal at normal operation ' CR assembly misalignment and drop
' CR ejection ' CR drop ' Boron dilution
' Loss of feedwater heating
&
&
&
(
&
&
&
(
&
&
&
&
(
&
Pressure abnormality
' Loss of external load ' Load increase
' Depressurization of core cooling system ' Isolation of main steam line
' Pressure control system failure
&
&
&
&
&
&
&
&
&
&
Reactor coolant flow abnormality
' Loss of offsite power
' Partial loss of reactor coolant flow ' Total loss of reactor coolant flow ' Total loss of feedwater flow ' Reactor coolant pump seizure ' Channel blockage
&
&
( (
&
&
(
&
(
&
&
( a) ( (c) Inadvertent
start or malfunction of core cooling
' Start-up of an inactive reactor coolant loop ' Inadvertent start-up of ECCS
' Inadvertent star-up of AFS
' Core coolant flow control system failure ' Feedwater control system failure
&
&
&
&
&
&
&
&
b) Abnormality in
secondary loop
' Depressurization ' Over SG water feed ' Loss of SG feedwater
&
&
&
Loss of reactor coolant
inventory
' LOCA
' Main steam line break ' SG tube rupture
' Main feedwater pipe rupture ' Generation of H2gas
' Dynamic load to containment
( ( ( ( (
( (
( (
( (
( ( Others ' Waste gas decay tank rupture
' Improper fuel assembly insertion or drop
( (
( (
( (
a) [3KP=HHKOOKBBAA@S=PANBHKS\ EOPDA O=IA =O[3KP=HHKOOKBNA=?PKN?KKH=JPBHKS\
b) [%AA@S=PAN?KJPNKHOUOPAI B=EHQNA\EOPDA O=IA =O[" KNA ?KKH=JPBHKS ?KJPNKHOUOPAI B=EHQNA\
c) Taken from sodium cooled reactors
Safety analysis of the two-flow pass Super FR with all upward flow also identified the accident of total loss of feedwater flow as important accident (Li and Oka, 2013). Flow rate due to coolant expansion also appeared in the second pass which mitigated the MCST before the coolant injection from AFS. However, after the coolant injection, the second pass of fuel channels experienced a high MCST due to operation of turbine control valves. Flow injection caused an increase of core pressure and in turn leaded the valves to open to higher position.
Afterwards the valves were quickly closed to lower position due to the output of the pressure controller. As a result the flow rate in the second pass was decreased leading to high MCST.
However, it was not happened in the first pass. In the two-flow pass core of Super FR characteristics of the first pass was driven by feedwater flow rate while the second pass was driven by turbine control valves operation as shown in Fig. 1-14. These characteristics were due to the large volume of lower, upper and gap plenums which separate the first and the second passes. The MCST was sensitive to the volume of the gap (Kamata and Oka, 2013). In the core design of the single-flow pass Super FR, however, the gap is removed. The characteristics of the accident might be different from that of the two-flow pass core. The water inventory is also less due to absent of the gap volume. It might also influence the safety characteristics of the single-flow pass core.
Fig. 1-14 Total loss of flow in the two-flow pass Super FR with all upward flow (Li and Oka, 2013)
1.3.3 Passive safety system of Super FR
Active safety systems have certain potential concerning termination of events or accidents that are effectively overcome with a protective system. However, it is limited by the reliability of the active safety system or prompt operator actions to prevent significant fuel failure and fission product release. Since the reliability of active systems cannot be decreased below a threshold and that of the operatorys action is debatable, there is growing concern about the safety of nuclear power plant due to the large uncertainty involved in probabilistic safety analysis particularly in analyzing human faults. In this view, a desirable goal for the safety characteristics of an innovative reactor is that its primary defense towards any serious accidents is achieved through its design features preventing the occurrence of such accidents without depending either on the operatorys action or the active system. That means the plant is facilitated with passive system to provide protection for any event that may lead to a serious accident (Abram and Elshahat, 2012).
Passive safety system of Super FR has not studied until now. However, some preliminary studies of passive system were conducted for supercritical light water cooled reactors (MacDonald et al., 2005; Wu et al., 2012). The passive safety system of Super FR is necessary to be studied.
1.4 Motivations and Objectives of Study
Start-up and safety analyses are important issues in core design development of a Super FR. This section summarizes the motivations and objectives of the current study which are based on the general and specific issues of start-up and safety characteristics of the current design of Super FR.
In general, since a Super FR is operated at supercritical pressure, the start-up system will be different from that of LWRs. Analysis of the star-up system is important because it involves transition of the coolant from subcritical pressure to supercritical pressure. There will be a large change of the coolant properties. There is a concern that the MCST might be high at around critical pressure during pressurization. The start-up system must assure that the MCST is low to keep the fuel rod integrity. Scheme of the pressurization from subcritical to supercritical pressure should be also investigated. Besides that, the start-up system should
provide superheated steam due to similar balance of plant (BOP) to that of fossil fuel power plants which the turbines are started by superheated steam. A start-up system with a recirculation had been proposed in the past study to meet these requirements. However, detailed thermal analysis of the start-up system was not carried yet. In the current study, the start-up system of the two-flow pass Super FR with all upward flow is analyzed in thermal view. The start-up procedures are investigated. Maximum cladding surface temperatures both during pressurization and power-raising are also clarified.
In view of the safety characteristics, since the operating pressure and cooling system are different from that of LWRs, the safety principle of a Super FR is also different. Comparison of the cooling system between LWRs and a Super FR is shown in Fig. 1-15. LWRs have coolant circulation system such as recirculation system of BWR and primary system of PWR.
LWRs have water level since these are operated at subcritical pressure. The fundamental safety requirement of LWR is keeping the coolant inventory by maintaining the water level in RPV of LWRs. A Super FR cooling system has no coolant circulation system and water level does not exist at supercritical pressure. The fundamental safety of a Super FR is keeping coolant flow rate instead of the water level. The safety analysis of a Super FR is important due to these differences.
Fig. 1-15 Cooling system of plant systems (Ishiwatari et al., 2005a)
Compared with thermal reactor (Super LWR), a Super FR has higher power density and no moderator. High power density means less water inventory in the RPV. Coolant flow rate
of a Super FR is lower than that of Super LWR. Thus the safety performance of a Super FR will be also different from that of Super LWR.
The single flow pass core of Super FR has some differences from the two flow pass core of Super FR. Gap and lower plenums of the two-flow pass core as shown in Fig. 1-7 are removed in the single-flow pass core design. It leads to be less water inventory in the single-flow pass core. It might influence the safety performance during abnormality. Besides that, a large change of coolant density in axial direction is introduced in the single-flow pass core of Super FR, leading to a high power peaking at beginning of cycle (BOC). In case of abnormality, it might also influence the safety performance. Large difference of flow distribution between Seed and Blanket assemblies leads to a large orifice coefficient in the blanket assembly. In case of flow rate recovery, high pressure drop may happen in the blanket assembly. MCST of the assembly should be clarified. These difference features lead to different safety characteristics of the single flow pass core from that of the two flow pass core of Super FR. Safety characteristics of the single flow pass core of Super FR are important to analyse.
Passive safety system of a Super FR has not studied until now. Passive safety system of a Super FR needs to be studied to improve the maintenance cost of safety system during operation. Passive safety systems do not need an emergency electric power. The current study proposes a concept of passive system for Super FR. Applicability of the passive system to a Super FR is investigated.
The objectives of the current study are as follow:
1. To perform start-up thermal analysis of the two-pass Super FR with all upward flow:
a. to investigate the startup procedures of the Super FR,
b. to perform the detailed time dependent start-up thermal analysis of the Super FR, c. to design the steam drum used for pressurization during the start-up.
2. To perform safety analysis of the single-flow pass core Super FR.
3. To propose a concept of passive safety system of Super FR 1.5 Thesis Organization
This thesis is divided into 6 chapters. Chapter 1 presents the background, core designs
description of the Super FR, review of the past studies, and the objectives of this study.
Chapter 2 presents the methods and codes that are used for the start-up and safety analyses.
Chapter 3 discusses the time dependent start-up thermal analysis of the two-flow pass Super FR with all upward flow. The start-up system and its procedures are clarified. Thermal analysis during pressurization especially at subcritical pressure which potential of high MCST may happen is discussed. Afterwards the thermal analysis during power-raising at supercritical pressure is presented.
Chapter 4 describes the safety analysis results of the single-flow pass Super FR. Its safety characteristics are clarified. Analysis results of abnormalities like accidents, abnormal transients, LOCA and ATWS are discussed. The analyses of accidents are carried both with and without considering the control system.
Chapter 5 presents the passive system of the Super FR. A Passive safety system is proposed for the Super FR. Applicability of the passive system to the Super FR is discussed.
Safety characteristics of the Super FR are clarified using the passive system.
Chapter 6 summaries the conclusions obtained from this study.
2. Analysis Method
2.1 Safety Analysis Codes
Two codes used for safety analysis of a Super FR were developed and validated at the University of Tokyo (Okano et al., 1996; Lee, 1996; Oka et al., 2010; Ikejiri et al., 2010; Li and Oka, 2013). The first code called as Supercritical Pressure Reactor Accident and Transient (SPRAT) was used for plant transient analysis of supercritical light water reactor. The main calculations of the code for both the start-up and safety analyses are mass and energy calculations, fuel rod heat transfer and point kinetics. The relation among them is shown in Fig. 2-1 (Oka et al., 2010). The second code is called as Supercritical-pressure Light Water Reactor LOCA Analysis (SCRELA) code which was used for LOCA analysis. In this study, firstly the calculation models of the start-up and safety analyses are developed based on the current core designs of the Super FRs. Afterwards the codes are modified based on the calculation models.
Fig. 2-1 Basic structure of SPRAT-F code (Oka et al., 2010)
2.2 Calculation Model
2.2.1 Calculation model of start-up analysis
Time dependent start-up thermal analysis of the two-pass core of Super FR with all upward flow is conducted. The start-up calculation model is developed based on the two-pass core design and the start-up system with recirculation. Calculation model of the start-up analysis is shown in Fig. 2-2 and the calculation flowchart is shown in Fig. 2-3.
Fig. 2-2 Calculation model for startup system analysis of the two-pass Super FR
Fig. 2-3 Flowchart of startup thermal analysis for the Super FR
A two-phase flow conservation equation is modeled in the start-up analysis code. The thermal hydraulic analysis is carried out by using the one-dimensional single-channel model
of the fuel assembly. The recirculation system is divided into eight parts: the main feedwater line, the lower dome, the reactor core (fuel channels), the gas plenums, the gap plenum, the upper plenum, the steam line and the steam drum. The feedwater and steam lines are divided into 10 nodes respectively. The lower dome and upper plenum are divided into 20 nodes respectively. The fuel channels and the gas plenums are divided into 36 and 18 nodes respectively. The gap is divided into 54 nodes and the steam drum is calculated by using 1 node. The heat exchanger is assumed to be an ideal heat exchanger. The thermo-dynamic properties of the coolant are assumed spatially uniform within each node.
The thermal calculation is carried out from the feedwater line to the steam drum. The inlet coolant temperature and flow rate are used as the boundary conditions. Coolant properties of density, temperature, enthalpy and mass flow rate in each node are calculated based on conservative laws of mass, energy and momentum, and a state equation which are included in the modified code.
2.2.2 Calculation model of safety analysis
2.2.2.1 Calculation model of accidents and abnormal transients
Fig. 2-4 Calculation model of safety analysis of the single-flow pass core Super FR
Safety analysis is applied for the single-flow pass core Super FR. The safety analysis is conducted both at supercritical and subcritical pressure. Calculation model of accident and abnormal transient analyses at supercritical pressure and during blowdown is shown in Fig.
2-4. The calculation flowchart is shown in Fig. 2-5. The model includes the reactor coolant pumps (RCPs), the safety system, and the fuel channels. The feedwater and the lower dome are divided into 10 and 20 nodes respectively. Each fuel channel is divided into 36 nodes axially. The gas and upper plenums are divided into 18 and 20 nodes respectively. The coolant properties are assumed spatially to be the same in each node.
Fig. 2-5 Flowchart of safety analysis code for the single-pass Super FR
The fuel channels are divided into average and hot channels. Axial power distribution of the average channel is taken from the average power of all fuel assemblies radially, while the power distribution of hot channel is taken from the hottest fuel rod where the maximum cladding temperature may happen. The hot channel is used to evaluate the MCST. Orifices are applied in the seed and blanket assemblies to set flow distribution at the average channel.
Orifices at the hot channel are used to set the MCST.
2.2.2.2 Calculation model of reflooding
Calculation model for reflooding phase at subcritical pressure after the blowdown is shown in Fig. 2-6. Flowchart of the calculation is shown in Fig. 2-7. The pumping head of LPCI, the injection momentum of ECCS water and the friction loss of liquid flow are neglected for the conservative calculation. The boundary conditions are water levels in the downcomer and the core, and pressures on these levels. The pressure on the downcomer water level is assumed to be the same as the containment pressure. The downcomer and lower dome are divided into 1 node respectively. The fuel channels (seed and blanket) are divided into 36 normally. The heat transfer coefficient sharply changes by about 2 orders of magnitude in the vicinity of a quench front. Therefore, to prevent the numerical instability caused by the abrupt change of the heat transfer, the neighboring nodes of the quench front are divided into 100 nodes (Lee, 1996).
Fig. 2-6 Calculation model for reflooding phase of the single-flow pass core Super FR
Fig. 2-7 Flowchart of reflooding calculation 2.3 Mass and Energy Conservation Models
The governing equations of mass and energy calculation are shown in Eqs. (2-1) and (2-2).
Mass conservation law:
Where ~: density (kg/m3) G: mass flow (kg/s) t: time (s)
z: position (m) h: enthalpy (J/kg) lf: mesh height (m)
Af: surface area of fuel pin (m2) Qx: heat flux (W/m2)
The discretization of the governing equations is done by an upstream scheme and a full implicit scheme. The discretized mass and energy equations are expressed by Eqs. (2-3) and (2-4).
Conditions of the inlet boundary are the feedwater flow rate both from reactor coolant pump (RCP) and auxiliary feedwater system (AFS), and the inlet temperature. Conditions of the outlet boundary are the flow rate through turbine control valve, turbine bypass valve, main steam isolation valves (MSIVs) and the safety relief valves (SRV). The characteristics of the turbine control valve, expressed as the change of steam flow rate, are shown in Fig. 2-8 (Ishiwatari et al. 2003). The feedwater flow rate changes with the core pressure as shown in Fig. 2-9 (Ishiwatari et al., 2003).
Fig. 2-9 Feedwater flow rate as function of core pressure (Ishiwatari et al., 2003)
2.4 Radial Heat Transfer Model
Heat transfer from fuel to coolant is calculated based on single channel model as shown in Fig. 2-10. The heat transfer is divided into 2 parts: heat transfer from fuel to cladding and heat transfer from cladding to coolant. The fuel pellet is divided into 4 nodes which the governing equation is as in Eq. (2-5).
Qproduct: Produced heat due to nuclear reaction (J) T: Temperature of fuel pellet (K)
V: Volume of fuel pellet of each mesh (m3)
~: Density of fuel pellet (kg/m3)
The produced heat by nuclear reaction, Qproduct, is evaluated by conversion factor, q0
which the produced heat is calculated as:
Qproduct= q0.n n: neutron number
Fig. 2-10 Model of radial heat transfer from fuel pellet to coolant (Ikejiri et al., 2010)
Heat transfer between the nodes of the fuel pellet is calculated by using conduction equation of Katachi-netsu as in Eq. (2-6).
Heat transfer from the fuel pellet to the cladding was calculated based on heat conduction through the gap between the fuel pellet and the cladding, and based on heat conduction through the cladding between the inside surface and outside surface of the cladding. The heat transfer through the gap was calculated by equation as in Eq. (2-8).
