ID: TECH / 2-1 Progress in design and engineering issues ...
1
Progress in design and engineering issues on JA DEMO Youji Someya 1 , Yoshiteru Sakamoto 1 , Ryoji Hiwatari 1 , Yuuya Miyoshi 1 , Satoshi Kakudate 1 , Hiroyasu Utoh 1 , Nobuyuki Asakura 1 , Yuki Homma 1 , Noriyoshi Nakajima 2 , Kenji Tobita 3 and the Joint Special Design Team for Fusion DEMO 1 National Institutes for Quantum and Radiological Science and Technology, 2 National Institute for Fusion Science, 3 Tohoku University [email protected] The paper presents solutions for critical problems in Japan’s DEMO (JA DEMO), which include common DEMO design issues beyond ITER-relevant technologies. The highlights of this design study are (1) Novel concept for water-cooled pebble bed blanket and tritium recovery, (2) System design for electric power generation with a management of the T concentration in the primary cooling system, (3) Identification of the classification for the rad-waste of JA DEMO, and (4) Identification of an accident sequences for safety guideline. The proposed concept as JA DEMO will be the foundation for Japan’s DEMO that can be envisioned in the next stage of ITER. INTRODUCTION DEVELOPMENT OF DEMO DESIGN ACTIVITY IN JAPAN In order to increase the feasibility of JA DEMO concept, studies on the following engineering issues were performed. 1. Countermeasure against a loss of coolant accident inside blanket (in-box LOCA) In the “honeycomb-shape”, little retention of tritium was found by flow analysis of purge gas. Simple concept of cylindrical structure are designed to meet the target TBR in the condition of the pressure tightness. 2. An outlook of the steady and stable power generation beyond several 100 MW The total power consumption was found to be 386 MW, and the electric output was evaluated to be 254 MWe Consistency between cooling system and T concentration control was confirmed for safety management. 3. Rad-waste management Even if uranium impurity in the beryllium as neutron multiplier was considered, all radwaste is classified as LLW and qualified for a shallow land burial. 4. Safety An accident sequences and mitigation systems are being sorted out. A mitigation systems of the countermeasures were confirmed on the safety analysis CONCLUSION ID: TECH / 2-1 [1] Y. Sakamoto et al., 27th IAEA Int. Conf. on Fusion Energy (2018) FIP/3-2 [2] Y. Someya et al., Fusion Eng. Des., 146 (2019), p. 894-897 [3] R. Hiwatari et al., Fusion Eng. Des. 143 (2019) 259-266 [4] K.Katayama et al., Fusion science tech.71(2017)261 [5] S. Tosti et al., Fusion Engin. Des. 43 (1998) 29–35 [6] J.H. Kim et al., J Nucl. Mater. 542 (2020) 152522 [7] Nakamura, et al., IEEE Trans. Plasma Sci. 44, 1689 (2016) REFERENCES Objective of DEMO Steady and stable power generation beyond several 100 MW Reasonable availability leading to commercialization Overall tritium breeding to fulfil self-sufficiency of fuels Application of as reliable technology as possible Design principle in the basic design phase Pre-conceptual design of JA DEMO Plasma operation Major radius, R P = ～ 8.5 m Arrangement of large CS coil for a few hour pulse operation in the commissioning phase Plasma perform., β N =3.4, HH=1.3 Study-state operation Fusion power P fus = 1.5 Allowable divertor heat load (Div. des. based on ITER technol.) Engineering technology ITER technol. as much as possible T breeding blanket: JA TBM strategy Divertor: Water cooling, W mono block Magnet: Radial Plate struc. CIC conductor (Nb 3 Sn) RECENT PROGRESS ON DESIGN ISSUES FOR JA DEMO 1. Countermeasure against a loss of coolant accident inside blanket (in-box LOCA) A conceptual design of blanket with pressure tightness against in-box LOCA has been carried out for safety of the JA DEMO. “square prism rib structure” is a reduction in the volume fraction of T breeder . [Ref.2] Overall TBR ≈ 1.0 (Target value≧1.05) Issue_1: T extraction is confirmed by CFD analysis with “honeycomb-shape” Square-shape Honeycomb-shape Circle-shape Honeycomb rib 〇 Cylindrical structure To thinner rib structure To simple structure Wasted space Issue_1: T extraction Issue_2: Complicated structure PF coil TF coil CS coil VV Divertor Blanket Issue_2: Simple concept of “cylindrical structure” is designed to meet the target TBR in the condition of the pressure tightness. • Target of TBR ( ≧ 1.05) was achievable with a honeycomb-rib • For the achievement of TBR target, P.F. to 80% with B.P. is necessary. • Issue: Amount of tritium retained in the breeding area may increase, due to the increase in pressure drop as a result of binary packing. The flow of He-purge gas was analyzed to confirm tritium retention 3 inflow points of He-purge gas are arranged near the FW Little retention of tritium in the area was found. Mole ratio of T Inflow of purge gas: 1 line Velocity : 5 m/s Pressure drop: 12.8 kPa T partial pressure: 17.2 Pa Inflow of purge gas: 3 line Velocity : 5 m/s Pressure drop: 16.0 kPa T partial pressure: 5.7 Pa Analysis results 1. Gap distribution is evaluated by DEM. 2. The flow analysis is performed in the distribution of binary packing area which is modelled as a porous body. Analysis method Cylindrical BB concept Calculation model Temp. distribution of BB DEMO construction Table DEMO Parameter [Ref.1] Steady state 2hrs pulse R p (m) / a p (m) 8.5 / 2.4 A 3.5 κ 95 1.65 q 95 4.1 I P (MA) 12.3 B T (T) 5.94 P fus ～ 1.5 ～ 1.0 Ave. NWL(MW/m 2 ) 1.0 0.7 Coolant water 290-325℃,15.5MPa Q 17.5 13 P ADD (MW) ～ 83.7 n e （10 19 m 3 ） 6.6 HH 98y2 1.31 1.13 β N 3.4 2.6 f BS 0.61 0.46 n e /n GW 1.2 2. Outlook of the steady and stable power generation beyond several 100 MW ● Cooling system PWR coolant water condition 15.5MPa, 290～325℃ Thermal efficiency：34.4% Fusion power 1500 MW Thermal power 1865 MW Generator output 640 MW Power consumption 386 MW Electric output 254 MW Breakdown of power balance H&CD power: 194 MW Freezing power: 91 MW Others: 30 MW Pumping power: 71 MW Thermal power: 1856 MW Generator output: 640 MW Electric output 254 MW Power generation systems is developed without inter. HEX. T permeation through the SG to the secondary cooling system in the PHTS is found to be less than the restricted amount of T disposal for PWR in Japan [Ref.3]. Issue_1: Evaluation of the power for DEMO plant consumption Issue_2: Consistency between cooling system and T concentration control Tritium impact in the primary coolant line T concentration at 1TBq/l was managed to apply the existing WDS for CANDU T permeation reduction factor is estimated to be 2077 from CANDU [Ref. 5] T permeation through a SG to the turbine system evaluated at 318 Ci/y/loop This value is less than restricted amount of T release for a PWR in JA [Ref.3] WDS Concentration of T 2.8×10 -3 gT/kg ( 1 TBq / kg from CANDU) 93.7 kg/hr Feed water into WDS 96 % Flow rate：7600 kg/s Extraction efficiency Tritium permeation Ref.4 0.24 gT/h (5.7 gT/day) FW：1.8 gT/day Breeder：2.3 gT/day DIV：1.6 gT/day Steam generator Turbine G Primary loop Secondary loop 318 Ci/year/loop (< T release for PWR in JA) 3. Rad-waste management for JA DEMO Disposal scenario Radioactive nuclide (RN) transport via likely pathways assessed based on JA regulation. All radwaste is classified as LLW and qualified for a shallow land burial. Using neutron multiplier is essential to ensure sufficient TBR. Amount of beryllide (Be 12 Ti) in DEMO is 500 t, which contain 6.8 kg of U. (※U in Be:20 wppm) The U content in Be needs to be less than approximately 0.85 wppm in order that the total radioactive concentration of α nuclides Np, U, Pu, Am and Cm produced from uranium is less than 10 10 Bq/ton of the total radioactive concentration for shallow land disposal along JA regulation. New purification process of QST has the ability to easily and stably remove U contained in Be until its content drops below allowable concentration (< 0.85 wppm) for shallow land disposal. T he concentration of U is removed to less than 0.1 ppm [Ref.6]. Calculation results Issue: U impurities in Be as N multiplier 4. Safety Previous safety study focus on the “bounding sequences” Lessons learned from “bounding sequences” However, in-vessel LOCA due to a large scale break of cooling pipe could result in a failure of VV (primary confinement boundary). Even for extremely hypothetical accidents, environmental release of tritium will be within a dose for evacuation-free. [Ref.7] Identification of an accident sequences and mitigation systems Mechanisms and countermeasures against threats are sorted out. Mechanisms Countermeasures In-VV LOCA After the in/ex-VV LOCA, LOOP... Air intrusion after In-VV LOCA Threats Loss of soundness of VV Overpressure in VV by coolant Overpressure in VV due to decay heat Overpressure in VV due to H explosion Installation of Suppression tank Installation of safety limiter Installation of decay heat removal system Adoption of N multiplier of low H production reaction rate Confirmed by the analysis. Installation of suppression tank on the VV Suppression tank is connected via the NBI port. Max. pressure at VV could be reduced to 15%. Tank of volume (water): 5,600 m 3 (2,800 m 3 ) Rupture disk: 0.2 MPa (Differential pressure) Cross section of the NB port: 4.2 m 2 A check valves are arranged to suppress Maximum pressure of VV is smaller than design pressure when the check valve arranges in the cooling system of the divertor baffle. Break area of the div. baffle is assumed to the all of the cooling pipe. Since the blanket has more amount of the water coolant than other IVCs, It can hardly contribute to suppress the max. pressure by check valve. Fig. Effect of a suppression tank on VV Fig. Effect of a check valve in cooling system for div. baffle External pressure load on VV Damage to VV due to flying objects, falling objects, etc. LOCA in Cryostat Short-circuit in TF coils, etc. Pipe whipping and jet during cooling pipe breaking Blanket module fall Installation of guard pipes Gap management between VV and TF coil Double walling of vacuum vessels and arrangement of fine reinforced ribs Installation of isolation valve, etc. Confirmation of countermeasures by analysis A simplified BB structure was proposed with a cylindrical structure . Characteristics of cylindrical concepts with Beryllide(Be 12 Ti) block. Be 12 Ti has little swelling compared to Be.Be 12 Ti can be used as blocks. Using blocks with a higher thermal conductivity than pebbles eliminates cooling piping inside the module. Allowable temp. of the materials is confirmed on the temp. distribution. All materials were within the allowable temperature range. The design is embodied with keeping the highest percentage of TBR values (Li 2 TiO 3 pebbles =25%). Dents around the Be 12 Ti block are designed to have the Li 2 TiO 3 ratio of 25% Target of overall TBR (>1.05) is achievable. Structure shape Casing (Out/In) :102.4 / 83.2 mm Number of cut: 16 Center direction: 10mm Width : 10.5mm Allowable temp. of material Be 12 Ti ≦ 900℃ Li 2 TiO 3 ≦ 900℃ F82H ≦ 550℃
Transcript of ID: TECH / 2-1 Progress in design and engineering issues ...
Progress in design and engineering issues on JA DEMOYouji Someya1, Yoshiteru Sakamoto1, Ryoji Hiwatari1, Yuuya Miyoshi1, Satoshi Kakudate1, Hiroyasu Utoh1, Nobuyuki Asakura1, Yuki Homma1,
Noriyoshi Nakajima2, Kenji Tobita3
and the Joint Special Design Team for Fusion DEMO1National Institutes for Quantum and Radiological Science and Technology,
2National Institute for Fusion Science, 3Tohoku [email protected]
The paper presents solutions for critical problems in Japan’s DEMO (JA DEMO),which include common DEMO design issues beyond ITER-relevant technologies. The highlights of this design study are (1) Novel concept for water-cooled pebble
bed blanket and tritium recovery, (2) System design for electric power generationwith a management of the T concentration in the primary cooling system,(3) Identification of the classification for the rad-waste of JA DEMO, and (4)Identification of an accident sequences for safety guideline.
The proposed concept as JA DEMO will be the foundation for Japan’s DEMOthat can be envisioned in the next stage of ITER.
INTRODUCTION DEVELOPMENT OF DEMO DESIGN ACTIVITY IN JAPAN
In order to increase the feasibility of JA DEMO concept,studies on the following engineering issues were performed.
1. Countermeasure against a loss of coolant accident inside blanket (in-box LOCA) In the “honeycomb-shape”, little retention of tritium was found by flow analysis of purge gas. Simple concept of cylindrical structure are designed to meet the target TBR in the condition
of the pressure tightness.2. An outlook of the steady and stable power generation beyond several 100 MW The total power consumption was found to be 386 MW, and the electric output was
evaluated to be 254 MWe Consistency between cooling system and T concentration control
was confirmed for safety management.3. Rad-waste management Even if uranium impurity in the beryllium as neutron multiplier was considered, all
radwaste is classified as LLW and qualified for a shallow land burial.4. Safety An accident sequences and mitigation systems are being sorted out. A mitigation systems of the countermeasures were confirmed on the safety analysis
CONCLUSION
ID: TECH / 2-1
[1] Y. Sakamoto et al., 27th IAEA Int. Conf. on Fusion Energy(2018) FIP/3-2
[2] Y. Someya et al., Fusion Eng. Des., 146 (2019), p. 894-897[3] R. Hiwatari et al., Fusion Eng. Des. 143 (2019) 259-266[4] K.Katayama et al., Fusion science tech.71(2017)261[5] S. Tosti et al., Fusion Engin. Des. 43 (1998) 29–35[6] J.H. Kim et al., J Nucl. Mater. 542 (2020) 152522[7] Nakamura, et al., IEEE Trans. Plasma Sci. 44, 1689 (2016)
REFERENCES
Objective of DEMO Steady and stable power generation beyond several 100 MW Reasonable availability leading to commercialization Overall tritium breeding to fulfil self-sufficiency of fuels
Application of as reliable technology as possible
Design principle in the basic design phase
Pre-conceptual design of JA DEMOPlasma operationMajor radius, RP = ～ 8.5 m
Arrangement of large CS coil for a few hour pulseoperation in the commissioning phase
Plasma perform., βN=3.4, HH=1.3Study-state operation
Fusion power Pfus = 1.5Allowable divertor heat load(Div. des. based on ITER technol.)
Engineering technology ITER technol. as much as possibleT breeding blanket: JA TBM strategyDivertor: Water cooling, W mono blockMagnet: Radial Plate struc. CIC conductor (Nb3Sn)
RECENT PROGRESS ON DESIGN ISSUES FOR JA DEMO1. Countermeasure against a loss of coolant
accident inside blanket (in-box LOCA)A conceptual design of blanket with pressure tightness against in-box LOCA has been carried out for safety of the JA DEMO.
“square prism rib structure” is a reductionin the volume fraction of T breeder. [Ref.2] Overall TBR ≈ 1.0 (Target value≧1.05)
Issue_1: T extraction is confirmed by CFD analysis with “honeycomb-shape”
Square-shape Honeycomb-shape
Circle-shape
Honeycomb rib
〇 Cylindrical structure
To thinner rib structure
To simple structureWasted space
Issue_1: T extractionIssue_2: Complicated
structure
PF coilTF coil CS coil
VVDivertor
Blanket
Issue_2: Simple concept of “cylindrical structure” is designed to meet the target TBR in the condition of the pressure tightness.
• Target of TBR ( ≧ 1.05) was achievable with a honeycomb-rib• For the achievement of TBR target, P.F. to 80% with B.P. is necessary.• Issue: Amount of tritium retained in the breeding area may increase,
due to the increase in pressure drop as a result of binary packing. The flow of He-purge gas was analyzed to confirm tritium retention
3 inflow points of He-purge gas are arranged near the FW Little retention of tritium in the area was found.
Mole ratio of T
Inflow of purge gas: 1 line Velocity : 5 m/s Pressure drop: 12.8 kPa T partial pressure: 17.2 Pa
Inflow of purge gas: 3 line Velocity : 5 m/s Pressure drop: 16.0 kPa T partial pressure: 5.7 Pa
Analysis results1. Gap distribution is evaluated by DEM.2. The flow analysis is performed
in the distribution of binary packing area which ismodelled as a porous body.
Analysis method
Cylindrical BB concept Calculation model Temp. distribution of BB
DEMO constructionTable DEMO Parameter [Ref.1]
Steady state 2hrs pulseRp (m) / ap (m) 8.5 / 2.4A 3.5κ95 1.65q95 4.1IP (MA) 12.3BT (T) 5.94Pfus ～ 1.5 ～ 1.0
Ave. NWL(MW/m2) 1.0 0.7Coolant water 290-325℃,15.5MPaQ 17.5 13PADD (MW) ～ 83.7ne（1019 m3） 6.6HH 98y2 1.31 1.13βN 3.4 2.6fBS 0.61 0.46ne/nGW 1.2
2. Outlook of the steady and stable power generation beyond several 100 MW
● Cooling system PWR coolant water condition 15.5MPa, 290～325℃
Thermal efficiency：34.4%
Fusion power 1500 MWThermal power 1865 MWGenerator output 640 MWPower consumption 386 MWElectric output 254 MW
Breakdown of power balance
H&CD power: 194 MW
Freezing power: 91 MW
Others: 30 MW
Pumping power: 71 MW
Thermal power: 1856 MW
Generator output: 640 MW
Electricoutput
254 MW
Power generation systems is developed without inter. HEX. T permeation through the SG to the secondary cooling
system in the PHTS is found to be less than the restrictedamount of T disposal for PWR in Japan [Ref.3].
Issue_1: Evaluation of the power for DEMO plant consumption
Issue_2: Consistency between cooling system and T concentration control
Tritium impact in the primary coolant line T concentration at 1TBq/l was managed to apply the existing WDS for CANDU T permeation reduction factor is estimated to be 2077 from CANDU [Ref. 5] T permeation through a SG to the turbine system evaluated at 318 Ci/y/loopThis value is less than restricted amount of T release for a PWR in JA [Ref.3]
WDS
Concentration of T2.8×10-3 gT/kg( 1 TBq / kg from CANDU)
93.7 kg/hrFeed water into WDS
96 %
Flow rate：7600 kg/s
Extraction efficiency
Tritium permeation Ref.40.24 gT/h(5.7 gT/day)
FW：1.8 gT/dayBreeder：2.3 gT/dayDIV：1.6 gT/day
Stea
mge
nera
tor
Turbine G
Primary loop Secondary loop
318 Ci/year/loop(< T release for PWR in JA)
3. Rad-waste management for JA DEMODisposal scenario Radioactive nuclide (RN) transport via likely
pathways assessed based on JA regulation. All radwaste is classified as LLW and qualified
for a shallow land burial.
Using neutron multiplier is essentialto ensure sufficient TBR.
Amount of beryllide (Be12Ti) in DEMO is 500 t,which contain 6.8 kg of U. (※U in Be:20 wppm)
The U content in Be needs to be less than approximately0.85 wppm in order that the total radioactive concentration of α nuclides
Np, U, Pu, Am and Cm produced from uranium is less than 1010 Bq/ton ofthe total radioactive concentration for shallow land disposal along JA regulation.
New purification process of QST has the ability to easily and stably remove U containedin Be until its content drops below allowable concentration (< 0.85 wppm) for shallowland disposal. The concentration of U is removed to less than 0.1 ppm [Ref.6].
Calculation results
Issue: U impurities in Be as N multiplier
4. SafetyPrevious safety study focus on the “bounding sequences” Lessons learned from “bounding sequences”
However, in-vessel LOCA due to a large scale break of cooling pipe could result in a failure of VV (primary confinement boundary).
Even for extremely hypothetical accidents, environmental release of tritium will be within a dose for evacuation-free. [Ref.7]
Identification of an accident sequences and mitigation systemsMechanisms and countermeasures against threats are sorted out.
MechanismsCountermeasures
In-VV LOCA
After the in/ex-VV LOCA, LOOP...
Air intrusion after In-VV LOCA
Threats Loss of soundness of VV
Overpressure in VV by coolant
Overpressure in VV due to decay heat
Overpressure in VV due to H explosion
Installation of Suppression tank
Installation of safety limiter
Installation of decay heat removal system
Adoption of N multiplier of low H production reaction rate
Confirmed by the analysis.
Installation of suppression tank on the VV Suppression tank is connected via the NBI port.Max. pressure at VV could be reduced to 15%.Tank of volume (water): 5,600 m3 (2,800 m3)Rupture disk: 0.2 MPa (Differential pressure)Cross section of the NB port: 4.2 m2
A check valves are arranged to suppressMaximum pressure of VV is smaller than design pressure when
the check valve arranges in the cooling system of the divertor baffle.Break area of the div. baffle is assumed to the all of the cooling pipe.
Since the blanket has more amount of the water coolant than other IVCs, It can hardly contribute to suppress the max. pressure by check valve.
Fig. Effect of a suppression tank on VV Fig. Effect of a check valve in cooling system for div. baffle
External pressure load on VV
Damage to VV due toflying objects, falling objects, etc.
LOCA in Cryostat Short-circuit in TF coils, etc.
Pipe whipping and jet during cooling pipe breaking
Blanket module fall
Installation of guard pipes
Gap management between VV and TF coil
Double walling of vacuum vessels and arrangement of fine reinforced ribs
Installation of isolation valve, etc.
Confirmation of countermeasures by analysis
A simplified BB structure was proposed with a cylindrical structure . Characteristics of cylindrical concepts with Beryllide(Be12Ti) block.Be12Ti has little swelling compared to Be.Be12Ti can be used as blocks.Using blocks with a higher thermal conductivity than pebbleseliminates cooling piping inside the module. Allowable temp. of the materials is confirmed on the temp. distribution. All materials were within the allowable temperature range. The design is embodied with keeping the highest
percentage of TBR values (Li2TiO3 pebbles =25%).Dents around the Be12Ti block are designedto have the Li2TiO3 ratio of 25% Target of overall TBR (>1.05) is achievable.
Structure shape Casing (Out/In)
:102.4 / 83.2 mm Number of cut: 16 Center direction: 10mm Width : 10.5mm
Allowable temp. of material Be12Ti ≦ 900℃ Li2TiO3 ≦ 900℃ F82H ≦ 550℃
