AEN WPRS Sodium Fast Reactor Core Definitions (version 1.2 September 19 th )

Transcription

1 AEN WPRS Sodium Fast Reactor Core Definitions (version 1.2 September 19 th ) David BLANCHET, Laurent BUIRON, Nicolas STAUFF CEA Cadarache laurent.buiron@cea.fr 1. Introduction and main objectives Taek Kyum KIM, Temitope TAIWO Argonne National Laboratory US-DOE The Generation IV International Forum (GIF) has defined the key research goals for advanced sodium-cooled fast reactors (SFR): - improved safety performance, specifically a demonstration of favourable transient behaviour under accident conditions; - improved economic competitiveness; - demonstration of flexible management of nuclear materials, in particular, waste reduction through minor actinide burning. With respect to SFR safety performance, one of the foremost GIF objective is to design cores that can passively avoid damage when the control rods fail to scram in response to postulated accident initiators (e.g., inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Under the auspices of the Working Party on Reactor and System (WPRS), a mandate has been proposed to work towards a shared analysis of the feedback and transient behaviour of next generation SFR concepts. In order to achieve these goals, a step-by-step analysis approach has been proposed: 1. Compile a state of the art report: review past and recent studies performed in the framework of sodium fast reactor and build a bibliographic repository which would stress core transient behaviours as a function of fuel characteristics (oxide, carbide, nitride and metal). 1

2 2. Perform a parametric study based on two different core sizes: large size core (3600 MW thermal) and medium size core ( MW thermal). For both cores sizes three types of fuel are proposed: oxide, carbide and metal. This comparative study is aimed at identifying the advantages and drawbacks for each concept based on nominal performances and global safety parameters: - Neutronics characterisation of global parameters (k-eff, power and flux distributions, void effect, Doppler, etc.) - Feed-back coefficient evaluation, discussion and agreement on corresponding calculation methodology. The expert group would provide initial core descriptions for both size cores. All the input data will be provided to the WPRS community and each contributor will perform step-by-step core analysis and benchmark comparisons using best estimate calculations. 3. Based on the results obtained in the previous step, transient calculations will be performed on a few selected cases for the principal unprotected transients (unprotected transient overpower (UTOP), unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS)) and the core behaviours characterised using a matrix classification. 4. Synthesis of the whole work into a final report including recommendations to improve safety and future work toward severe accidents and minor actinides management. The step 2 above is aimed at providing core descriptions to be used for neutronics performance as well as feedback coefficient evaluations and further transient calculations. In the present document, two core designs are currently being proposed for the 3600 MWth Sodium-cooled Fast Reactor (SFR) concept: one is based on oxide fuel and the other on carbide fuel. For medium size reactors, descriptions are provided for two 1000MWt ABR cores using metallic and oxide fuels, respectively. The following data are mainly geometry and materials definitions that enable neutronics characterisation of the cores for both equilibrium beginning and end of cycle states. At this stage, no thermal-hydraulic data are provided. 2. Core descriptions 2.1. Large size cores Core descriptions have been provided by CEA for two large-core 3600 MWth concepts. Both have medium power densities that result in low reactivity swing during the equilibrium burn cycle. Both concepts use Oxide Strengthened Steel (ODS) cladding with helium bond. The oxide core is based on the fat pin with small wire concept that enables to reach selfbreeding without fertile blanket. The resulting core exhibits an average burnup around 100 GWd/tHM for a corresponding cycle length of 410 equivalent full power days with one fifth reloading scheme. 2

3 The carbide core was designed to fit a very low linear rate to give an enhanced margin to fuel melting. The core exhibits an average burnup close to 70 GWd/t determined by the fuelcladding mechanical interaction limit. As both cores have a low reactivity swing, they share the same basic design (pin definition) for primary and secondary control systems. The operation conditions are specified in Table 1. Reactor power (MWth) 3600 Core inlet temperature ( C) 395 Core outlet temperature ( C) 545 Average core structure temperature ( C) 470 (structure, absorber and coolant medium) Average fuel temperature ( C) 1227 (oxide core) / 987 (carbide core) Carbide Core Table 1: Nominal conditions The 3600 MWth SFR carbide core layout is presented in Figure 1. The core consists of 487 fuel, 270 radial reflector and 27 control subassemblies. The core is divided into inner and outer core zones, which are composed of 286 and 201 fuel assemblies, respectively. Two independent safety-grade reactivity control sub-systems are used. The primary control system consists of 6 control subassemblies in the inner core and 12 control subassemblies at the interface between the inner and the outer zone. The secondary system contains 9 control subassemblies located between the 7 th and 8 th rows. Although the core is surrounded by various materials, a vacuum boundary condition (i.e., no-return current) has been specified for neutronics modelling. Figure 1 : Radial Core Layout of the 3600 MWth SFR, carbide fuel 3

4 The fuel sub-assembly consists of a hexagonal wrapper tube that contains a triangular arrangement of helium bonded fuel pins with helical wire wrap spacers. The hexagonal wrapper tube and the wire wrap spacers are made of EM10-like steel. The volume of wire wrap spacers is included in the cladding volume by means of radius increase in order to simplify the pin description. The fuel pin consists of (U,Pu)C pellets with Oxide Strengthened Steel (ODS) cladding. The fuel density is smeared to account for swelling during irradiation. In the present case, it is assumed that all fuel slugs are in contact with the cladding at the beginning of cycle for simplicity. The fuel sub-assembly characteristics are summarized in Table 2. All the values are given at operating conditions. Overall length of subassembly - Lower Gas Plenum - Lower Axial reflector - Active core height - Upper Gas plenum - Upper Axial Reflector Unit Operating state cm Subassembly pitch, cm cm Subassembly duct outer flat-to-flat distance cm Subassembly duct wall thickness cm Number of fuel pins 469 Outer radius of cladding cm a) Inner radius of cladding cm Fuel slug radius cm Pin to Pin distance cm a) Cladding outer radius is increased to compensate for the smearing of the wire wrap. Table 2: Parameters of the fuel sub-assembly for the carbide core The axial description of the fuel subassembly is presented in Figure 2. The axial pin design is based on a central 1 meter active zone surrounded by two gas plenum. The upper one accounts for the top of the pin and has a limited dimension. The axial reflector at the bottom of the active zone is composed of steel pellets located in the pin. The same composition is used also for upper axial reflector for simplicity. The volume fraction of each axial part of the fuel subassembly is presented in Table 3. 4

5 Figure 2 : Schematic axial description of the fuel subassembly, carbide fuel Fuel Sodium ODS EM10 Fuel Axial Reflector Upper gas Plenum Lower Gas Plenum Table 3 : Volume fraction of fuel sub-assembly for the carbide core Oxide Core The 3600 MWth SFR oxide core layout is presented in Figure 3. The core consists of 453 fuel, 270 radial reflector and 27 control subassemblies. The core is divided into inner and outer core zones, which are composed of 225 and 228 fuel assemblies, respectively. Two independent safety-grade reactivity control sub-systems are used. The primary control system consists of 6 control subassemblies in the inner core and 12 control subassemblies at the interface between the inner and the outer zones. The secondary system contains 9 control subassemblies located in the 7 th row. Although the core is surrounded by various materials, a vacuum boundary condition (i.e., no-return current) is specified for neutronics modelling. 5

6 Figure 3: Radial Core Layout of the 3600 MWth SFR core, oxide fuel The fuel sub-assembly consists of a hexagonal wrapper tube that contains a triangular arrangement of helium bonded fuel pins with helical wire wrap spacers. The hexagonal wrapper tube and the wire wrap spacers are made of EM10-like steel. The volume of wire wrap spacers is included in the cladding volume by means of radius increase in order to simplify the pin description. The fuel pin consists of (U,Pu)O 2 pellets with Oxide Strengthened Steel (ODS) cladding. The fuel sub-assembly characteristics are summarized in Table 4. All the values are given at operating conditions. Overall length of subassembly - Lower Gas Plenum - Lower Axial reflector - Active core height - Upper Gas plenum - Upper Axial Reflector Unit Operating state cm Subassembly pitch cm Subassembly duct outer flat-to-flat distance cm Subassembly duct wall thickness cm Number of fuel pins 271 Outer radius of cladding cm a) Inner radius of cladding cm Fuel slug radius cm Inner central hole radius (helium) cm Pin to Pin distance cm b) Cladding outer radius is increased to compensate for the smearing of the wire wrap. 6

7 Table 4: Parameters of the fuel sub-assembly for the oxide core Figure 4 : Schematic axial description of the fuel subassembly, oxide fuel The axial description of the fuel subassembly is presented in Figure 4 (the inner central hole, treated like a smearing factor on fuel concentrations, is not displayed here). The axial pin design is based on a central 1 meter active zone surrounded by two gas plenum. The upper one accounts for the top of the pin and has a limited dimension. The axial reflector at the bottom of the active zone is composed of steel pellets located in the pin. The same composition is used also for upper axial reflector for simplicity. The volume fraction of each axial part of the fuel subassembly is presented in Table 5. Fuel Sodium ODS EM10 Fuel Axial Reflector Upper gas Plenum Lower Gas Plenum Table 5 : Volume fraction (%) of fuel sub-assembly for the oxide core 7

8 Control Rod and Radial Reflector Design Both the 3600 MWth carbide- and oxide-fuel cores use the same control rod description for primary and secondary control systems. It consists of hexagonal lattice of sodium bonded boron carbide pins of wire wrap spacers inside several ducts. The volume of wire wrap spacers is included in the cladding volume by means of radius increase in order to simplify the pin description. Due to a low reactivity swing, the primary system uses natural boron carbide, while the secondary system uses enriched B10 boron carbide. Duct and Cladding structure used the EM10 material. The Primary Control sub-assembly characteristics are summarized in Table 6 and illustrated in Figure 5. Unit Carbide core Oxide core Subassembly pitch cm Sodium gap width inter assembly cm Subassembly duct flat-to-flat width cm Wrapper tube thickness cm Outer flat-to-flat internal duct width cm Inner flat-to-flat internal duct width cm Number of pins cm Outer cladding diameter cm Inner cladding diameter cm Pellet diameter cm Pellet material B 4 C (natural) B 4 C (natural) Pin to Pin distance cm Table 6: Primary Control Rod characteristics 8

9 / / / Primary control L = Legend: B 4 C EM10 EM10 Na Na Figure 5: Primary control subassembly description (carbide and oxide core) The Secondary Control subassembly characteristics are summarized in Table 7 and illustrated in Figure 6. Unit Carbide core Oxide core Subassembly pitch cm Sodium gap width inter assembly cm Subassembly duct flat-to-flat width cm Wrapper tube thickness cm Internal duct outer diameter cm Internal duct inner diameter cm Number of pins cm Outer cladding diameter cm Inner cladding diameter cm Pellet diameter cm Pellet material B 4 C (90% B10) B 4 C (90% B10) Pin to Pin distance cm Table 7: Secondary Control Rod characteristics 9

10 / / / Secondary Control L = Φ / Legend : B 4 C EM10 EM10 Na Na Figure 6: Secondary control subassembly description (carbide and oxide core) Primary Control Secondary Control Empty Duct Carbide Oxide Carbide Oxide Carbide/Oxide B4C Coolant Structure (EM10) Table 8 : Volume fraction of primary and secondary subassembly (%) For both control rod systems, the absorber height is the same as the active zone ( cm). For this benchmark, they remain located at the top of the active core zone (interface between active core zone and upper gas plenum). The other part of the sub-assembly consists of an empty duct filled with sodium. 10

11 For radial reflector, a single unique homogeneous medium is used for both oxide and carbide core. This medium spread along the overall length of the corresponding sub-assembly. The associated volume fractions are 26% for sodium and 74% for steel (EM10 material) Material Description Fuel Materials Data for the nominal operating condition are presented and were calculated by accounting for the effects of thermal expansion and irradiation swelling from the fuel fabrication state. The homogenized compositions of fuel sub-assembly are given for both beginning and end of cycle for each core for hot conditions. Here, the subassembly is divided into five axial concentration sets for each different initial Pu content (inner core and outer core). Fission product isotopes were replaced by a representative isotope (Mo) in terms of equivalent absorption, and only one averaged value is available for each active zone (inner and outer). The averaged pin compositions for the zones are given for hot condition from Table 9 to Table 16. For each core, both Beginning and End of Equilibrium Cycle (BOEC and EOEC) number densities are given. Number densities lower than atoms/barn have been omitted. Here, Am242g stands for ground state. Carbide Core Upper boundary from active core bottom (cm) Nuclide C E E E E E-02 U E E E E E-06 U E E E E E-05 U E E E E E-06 U E E E E E-02 Np E E E E E-06 Np E E E E E-06 Pu E E E E E-04 Pu E E E E E-03 Pu E E E E E-03 Pu E E E E E-04 Pu E E E E E-04 Am E E E E E-05 Am242g E E E E E-09 Am242m E E E E E-07 Am E E E E E-05 Cm E E E E E-09 Cm E E E E E-06 Cm E E E E E-08 Cm E E E E E-07 11

12 Cm E E E E E-08 Cm E E E E-10 Mo E E E E E-03 Table 9 : Number Densities of Inner Core Fuel Pin, Carbide Core, BOEC (atoms/barn-cm) Upper boundary from active core bottom (cm) Nuclide C E E E E E-02 U E E E E E-06 U E E E E E-05 U E E E E E-06 U E E E E E-02 Np E E E E E-06 Np E E E E E-06 Pu E E E E E-04 Pu E E E E E-03 Pu E E E E E-03 Pu E E E E E-04 Pu E E E E E-04 Am E E E E E-05 Am242g E E E E E-08 Am242m E E E E E-06 Am E E E E E-05 Cm E E E E E-06 Cm E E E E E-08 Cm E E E E E-06 Cm E E E E E-08 Cm E E E E E-09 Cm247 Mo E E E E E-03 Table 10 : Number Densities of Outer Core Fuel Pin, Carbide Core, BOEC (atoms/barn-cm) 12

13 Oxide Core Upper boundary from active core bottom (cm) Nuclide O E E E E E-02 U E E E E E-06 U E E E E E-05 U E E E E E-06 U E E E E E-02 Np E E E E E-06 Np E E E E E-06 Pu E E E E E-05 Pu E E E E E-03 Pu E E E E E-03 Pu E E E E E-04 Pu E E E E E-04 Am E E E E E-05 Am242g E E E E E-08 Am242m E E E E E-06 Am E E E E E-05 Cm E E E E E-06 Cm E E E E E-08 Cm E E E E E-06 Cm E E E E E-08 Cm E E E E E-09 Cm247 Mo E E E E E-03 Table 11 : Number Densities of Inner Core Fuel Pin, Oxide Core, BOEC (atoms/barn-cm) 13

14 Upper boundary from active core bottom (cm) Nuclide O E E E E E-02 U E E E E E-06 U E E E E E-05 U E E E E E-06 U E E E E E-02 Np E E E E E-06 Np E E E E E-06 Pu E E E E E-04 Pu E E E E E-03 Pu E E E E E-03 Pu E E E E E-04 Pu E E E E E-04 Am E E E E E-05 Am242g E E E E E-08 Am242m E E E E E-06 Am E E E E E-05 Cm E E E E E-06 Cm E E E E E-08 Cm E E E E E-06 Cm E E E E E-08 Cm E E E E E-09 Cm247 Mo E E E E E-03 Table 12 : Number Densities of Outer Core Fuel Pin, Oxide Core, BOEC (atoms/barn-cm) 14

15 Structure, Coolant and Absorber materials Both cores used the same cladding, duct and absorber materials composition. Table 13 and Table 14 present the number densities at nominal operation condition to be used in this benchmark. Element Duct (EM10) Cladding (ODS) Na C E E-04 O E-04 Si E-04 Ti E E-04 Cr E E-02 Fe E E-02 Ni E E-04 Mo E-04 Mn E E-04 P E-05 Al E-03 Co E-04 Cu E-04 Y E-04 Coolant (Na) E-02 Table 13 : Structure and Coolant Material Number Densities (atoms/barn) Element Primary Control Secondary Control C 2.70E E-02 B E E-02 B E E-03 Table 14 : Absorber Material Number Densities (atoms/barn) 2.2. Medium size cores The numerical benchmark specifications based on the reference 1000 MWt Advanced Burner Reactor (ABR) metallic and oxide core concepts [1] are presented. The ABR core concepts were developed for the study of future fast reactor design options under the Global Nuclear Energy Partnership (GNEP) program. Compact core concepts with a transuranics (TRU) conversion ratio of ~0.7 were developed for a one-year cycle length with 90% capacity 15

16 factor. Conventional or reasonably proven materials were utilized in the ABR core concepts so that the core stays within current fast reactor technology knowledge base. Data are provided for the nominal operating condition of the ABR cores using fuel compositions associated with the equilibrium cycle. The ABR equilibrium cycle was determined by recycling all discharged TRU with an external makeup TRU feed recovered from LWR used nuclear fuel Metallic core Figure 7 shows the radial core layout of the 1000 MWt ABR metallic benchmark core. The core consists of 180 drivers, 114 radial reflectors, 66 radial shields, and 19 control subassemblies. The core is divided into inner and outer core zones, which are composed of 78 and 102 driver assemblies, respectively. Two independent safety-grade reactivity control sub-systems are used. The primary control system consists of three control subassemblies in the fourth row and 12 control subassemblies in the seventh row. The secondary system contains four control subassemblies located at the core center and in the fourth row. Although the core is surrounded by various materials, a vacuum boundary condition (i.e., noreturn current) is imposed in the benchmark. Figure 7 : Radial Core Layout of ABR Metallic-Fuel Core The nominal power is 1000 MWt and the core inlet and bulk outlet temperatures are 355 C and 510 C, respectively. The core temperature distribution is dependent on the power distribution, but uniform temperature values in the coolant, structural materials, and fuel have been assumed in this benchmark problem for simplicity. Table 15 contains data for the nominal operating condition. 16

17 Unit Value Reactor Power MW-thermal Coolant temperature C Average core structural temperature C Average metallic fuel temperature C Table 15 : Nominal Operating Condition, Metallic Core The design parameters of the driver, radial reflector, radial shield, and control subassemblies are provided in Table 16 to Table 19, respectively. Data for the nominal operating condition are presented and were calculated by accounting for the effects of thermal expansion and irradiation swelling from the fuel fabrication state. At the fabrication state, the fuel pin and control rod are helically wrapped with wire for accommodating the coolant flow. Here, the wire-wrap has been smeared with the cladding in order to simplify the cladding geometry. As a result, the outer radii of the fuel and control rod claddings are slightly increased to compensate for the smeared wire-wrap. For accommodating irradiation induced swelling, the smeared density of the fresh fuel is 75% and bond sodium is used to fill the gap between the metallic fuel slug and cladding. Although fresh and burned fuels co-exist at the beginning of the equilibrium cycle (BOEC), it is assumed that all fuel slugs contact the cladding, the fuel slug grows 5% by the effect of irradiation swelling, and the bond sodium is displaced to the lower part of the upper gas plenum. Figure 8 to Figure 11 show the schematics of the driver, radial reflector, radial shield, and control subassemblies, respectively, and the volume fractions at the nominal operating condition are provided in Table 20. For all subassemblies, the duct and cladding are made of HT-9. Subassemblies characteristics are described from Table 16 to Table 19. Each driver subassembly contains 271 fuel pins arranged in a triangular pitch array. The fuel and coolant volume fractions are 39% and 35%, respectively. The sodium volume fraction in the lower part of the gas plenum increases to 74% because of the displaced bond sodium and the helium gas in the gas plenum is ignored. The lower structure is assumed to be a homogeneous mixture of sodium (70%) and SS-316 (30%), and the lower reflector consists of 271 solid HT-9 pins. For simplicity, the upper structure is assumed to be identical to the lower reflector. The reflector subassembly contains 91 solid HT-9 pins arranged in a triangular pitch array. The shield subassembly consists of 19 thick HT-9 tubes (cladding) containing boron carbide pellets. Natural boron (19.1% atomic fraction) is used with 81% smeared B 4 C pellet density. The control subassembly consists of two ducts: interior and outer ducts. The outer duct is identical to the driver subassembly duct. The external dimension of the interior duct is smaller than that of the outer duct to allow free motion within the outer duct. Seven control rods are contained in the interior duct. The control rod consists of a HT-9 tube containing boron carbide pellets. Enriched boron (65% atomic fraction) is used with 85% smeared B 4 C pellet density. For simplicity of the depletion model, all structural materials including boron are assumed to be non-depleting. In this benchmark, it is assumed that the lower structure, 17

18 lower reflector and upper structure of the reflector, shield and control subassemblies are identical to those of the driver subassembly. Unit Operating state Overall length of subassembly cm Lower structure Lower reflector Active core height Replaced bond sodium Gas plenum Upper structure Subassembly pitch, cm cm Subassembly duct outer flat-to-flat distance cm Subassembly duct wall thickness cm Number of fuel pins 271 Outer radius of cladding cm a) Inner radius of cladding cm Fuel slug radius cm Pin to Pin distance cm a) Cladding outer radius has been slightly increased to compensate for the smearing of the wire wrap with the cladding. Table 16 : Parameters for Driver Subassembly of ABR Metallic Core Overall length of subassembly - Lower structure - Lower reflector - Radial reflector - Upper structure Unit Operating state cm Subassembly pitch, cm cm Subassembly duct outer flat-to-flat distance cm Subassembly duct wall thickness cm Number of fuel pins 91 Rod radius cm Pin to Pin distance cm Table 17 : Parameters for Radial Reflector Subassembly of ABR Metallic Core 18

19 Overall length of subassembly - Lower structure - Lower reflector - Radial shield - Upper structure Unit Operating state cm Subassembly pitch, cm cm Subassembly duct outer flat-to-flat distance cm Subassembly duct wall thickness cm Number of fuel pins 19 Outer radius of cladding cm Inner radius of cladding cm Absorber radius cm Pin to Pin distance cm Table 18 : Parameters for Shielding Subassembly of ABR Metallic Core Overall length of subassembly - Lower structure - Lower reflector - Absorber - Empty duct Unit Operating state cm Subassembly pitch, cm cm Subassembly duct outer flat-to-flat distance cm Subassembly duct wall thickness cm Interior duct outer flat-to-flat distance cm Interior duct wall thickness cm Number of fuel pins 7 Outer radius of cladding cm a) Inner radius of cladding cm Absorber radius cm Pin to Pin distance cm a) Cladding outer radius is increased to compensate for the smearing of the wire wrap. Table 19 : Parameters for Control Assembly of ABR Metallic Core 19

20 Fuel Upper structure cm Ø mm Ø mm Gas plenum cm He Ø mm Ø mm A A Replace sodium cm Active core cm Bond sodium Ø mm Ø mm cm Ø mm Lower reflector cm HT-9 Section A-A Lower structure cm Fuel pin Duct Figure 8 : Schematics of Driver Subassembly of ABR Metallic Core 20

21 Lower structure cm Lower reflector cm Radial reflector cm cm Upper structure cm Figure 9 : Schematics of Radial Reflector Subassembly of ABR Metallic Core 21

22 Ø mm Shield rod cm Shield Ø mm Lower structure cm cm Lower reflector cm Upper structure cm A A Section A-A Shield rod Duct Figure 10 : Schematics of Radial Shield Subassembly of ABR Metallic Core 22

23 Empty duct cm Subassembly duct Interior duct Control absorber cm A A Absorber Ø mm Ø mm Empty duct cm (active core) Lower structure cm Lower reflector cm Section A-A Control Rod Assembly (when fully withdrawn) Figure 11 : Schematics of Control Subassembly of ABR Metallic Core 23

24 Region Coolant HT-9 Fuel Natural Enriched B 4 C B 4 C Lower structure Homogeneous mixture of 30% SS-316 and 70% Sodium Lower reflector Upper structure Active core Driver Displaced bond sodium Gas plenum Radial reflector Radial shield Absorber in control subassembly Empty duct in control subassembly Table 20 : Volume Fractions of ABR Metallic Core (%) Material Description for Metallic Core The fuel compositions are obtained from the reference ABR core data at the BOEC. For the reference ABR core concept, the equilibrium cycle was determined by recycling all discharged TRU. For external makeup feed, TRU recovered from LWR used nuclear fuel is assumed. At the equilibrium cycle, the TRU conversion ratio is ~0.7 and cycle length is days with full power operation. Material Nuclide Number density Na E-02 Lower structure Fe E-02 (homogeneous Ni E-03 mixture of SS-316 Cr E-03 and Sodium) Mn E-04 Mo E-04 Coolant Na E-02 Fe E-02 Ni E-04 HT-9 Cr E-02 Mn E-04 Mo E-04 C E-02 Natural B 4 C B E-02 B E-02 C E-02 Enriched B 4 C B E-02 B E-02 Table 21 : Number Densities of Coolant and Structural Materials of ABR Metallic Core (atoms/barn-cm) 24

25 As for large core descriptions, the variation of fuel composition in the irradiated core is simplified. The active core is axially divided into five zones and uniform fuel composition is assumed in each zone. Thus, ten different fuel compositions are provided in this benchmark: five each for the inner and outer cores, respectively. A simplified fission products model is also adopted with a pseudo fission product. Natural molybdenum is used to represent the fission products (pseudo fission product). Table 21 provides the number densities of coolant and structural materials. The fuel pin compositions for the inner and outer cores at BOEC are provided in Table 22 and Table 23, respectively. All data are obtained for the nominal operating condition by adjusting for the effects of thermal expansion and irradiation swelling. As mentioned in previous sections, it is assumed that all subassemblies have the same lower structure, lower reflector and upper structure and the material volume fractions are provided in Table 20. It is noted that the homogenized number densities at the nominal operating condition can be calculated by using the volume fractions of Table 20 and the number densities provided in Table 21 to Table 23 Nuclide Upper boundary from active core bottom (cm) U E E E E E-06 U E E E E E-05 U E E E E E-06 U E E E E E-02 Np E E E E E-05 Pu E E E E E-10 Pu E E E E E-04 Pu E E E E E-03 Pu E E E E E-03 Pu E E E E E-04 Pu E E E E E-04 Am E E E E E-04 Am-242 m E E E E E-06 Am E E E E E-04 Cm E E E E E-06 Cm E E E E E-07 Cm E E E E E-05 Cm E E E E E-05 Cm E E E E E-06 Zr E E E E E-03 a) Mo E E E E E-04 a) Representative for fission products Table 22 : Number Densities of Inner Core Fuel Pin, Metallic Core, BOEC (atoms/barn-cm) 25

26 Nuclide Upper boundary from active core bottom (cm) U E E E E E-06 U E E E E E-05 U E E E E E-06 U E E E E E-02 Np E E E E E-04 Pu E E E E E-10 Pu E E E E E-04 Pu E E E E E-03 Pu E E E E E-03 Pu E E E E E-04 Pu E E E E E-04 Am E E E E E-04 Am-242 m E E E E E-05 Am E E E E E-04 Cm E E E E E-06 Cm E E E E E-07 Cm E E E E E-05 Cm E E E E E-05 Cm E E E E E-05 Zr E E E E E-03 a) Mo E E E E E-04 a) Representative for fission products Table 23 : Number Densities of Outer Core Fuel Pin, Metallic Core, BOEC (atoms/barn-cm) Oxide core The ABR oxide core concept was developed to allow the interchange of metal and oxide fuel sub-assemblies. Thus, the total number of driver sub-assemblies, the locations of the control sub-assemblies, and outer-dimensions of the oxide-fuel sub-assembly are identical to those of the metallic core. However, the internal sub-assembly design parameters such as fuel pin diameter, volume fractions, active fuel height, etc., and the arrangement of the driver subassemblies were determined to meet the ABR design goal of a compact core concept with medium TRU conversion ratio and one-year operation. Figure 12 shows the radial layout of the 1000 MWt ABR oxide benchmark core. The core consists of 180 drivers, 114 radial reflectors, 66 radial shields, and 19 control subassemblies. The core is divided into inner, middle, and outer core zones, which are composed of 30, 90, and 92 driver assemblies, respectively. A vacuum boundary condition is also imposed in the ABR oxide benchmark core. 26

27 Figure 12 : Radial Core Layout of ABR Oxide-Fuel Core The nominal power and temperatures of the oxide core are identical to those of the metallic core (see Table 15) except for the average fuel temperature. The average fuel temperature of the oxide fuel is assumed to be 1027 C. The design parameters of the driver, radial reflector, radial shield, and control subassemblies are provided in Table 24 to Table 27, respectively. Data for the nominal operating condition are presented and were calculated by accounting for the effects of thermal expansion and irradiation swelling from the fuel fabrication state. At the fabrication state, the fuel pin and control rod are helically wrapped with wire for accommodating the coolant flow, but the wire-wrap has been smeared with the cladding in order to simplify the cladding geometry. Irradiation induced swelling of the oxide fuel is ignored and it is assumed that fuel pellet contacts the cladding with 85% theoretical density oxide fuel. Figure 13 to Figure 16 show the schematics of the driver, radial reflector, radial shield, and control subassemblies, respectively, and the volume fractions at the nominal operating condition are provided in Table 28. For all subassemblies, the duct and cladding is made of HT-9. 27

28 Overall length of subassembly - Lower structure - Lower reflector - Active core height - Gas plenum - Upper structure Unit Operating state cm Subassembly pitch, cm cm Subassembly duct outer flat-to-flat distance cm Subassembly duct wall thickness cm Number of fuel pins 271 Outer radius of cladding cm a) Inner radius of cladding cm Fuel pellet radius cm a) Cladding outer radius has been slightly increased to compensate for the smearing of the wire wrap with the cladding. Table 24 : Parameters for Driver Subassembly of ABR Oxide Core Overall length of subassembly - Lower structure - Lower reflector - Radial reflector - Upper structure Unit Operating state cm Subassembly pitch, cm cm Subassembly duct outer flat-to-flat distance cm Subassembly duct wall thickness cm Number of fuel pins 91 Rod radius cm Table 25 : Parameters for Radial Reflector Subassembly of ABR Oxide Core Overall length of subassembly - Lower structure - Lower reflector - Radial shield - Upper structure Unit Operating state cm Subassembly pitch, cm cm Subassembly duct outer flat-to-flat distance cm Subassembly duct wall thickness cm Number of fuel pins 19 Outer radius of cladding cm Inner radius of cladding cm Absorber radius cm Table 26 : Parameters for Shielding Subassembly of ABR Oxide Core 28

29 Overall length of subassembly - Lower structure - Lower reflector - Absorber - Empty duct Unit Operating state cm Subassembly pitch, cm cm Subassembly duct outer flat-to-flat distance cm Subassembly duct wall thickness cm Interior duct outer flat-to-flat distance cm Interior duct wall thickness cm Number of fuel pins 7 Outer radius of cladding cm a) Inner radius of cladding cm Absorber radius cm a) Cladding outer radius is increased to compensate for the smearing of the wire wrap. Table 27 : Parameters for Control Assembly of ABR Oxide Core 29

30 Figure 13 : Schematics of Driver Subassembly of ABR Oxide Core 30

31 Figure 14 : Schematics of Radial Reflector Subassembly of ABR Oxide Core 31

32 Figure 15 : Schematics of Radial Shield Subassembly of ABR Oxide Core 32

33 Figure 16 : Schematics of Control Subassembly of ABR Oxide Core 33

34 Region Coolant HT-9 Fuel Natural Enriched B 4 C B 4 C Lower structure Homogeneous mixture of 30% SS-316 and 70% Sodium Lower reflector Upper structure Driver Active core Gas plenum Radial reflector Radial shield Absorber in control subassembly Empty duct in control subassembly Table 28 : Volume Fractions of ABR Oxide Core (%) Each driver subassembly contains 271 fuel pins arranged in a triangular pitch array. The fuel and coolant volume fractions are 41% and 33%, respectively. The lower structure is assumed to be a homogeneous mixture of sodium (70%) and SS-316 (30%), and the lower reflector consists of 271 solid HT-9 pins. For simplicity, the upper structure is assumed to be identical to the lower reflector. The design parameters of the reflector, shield, and control subassemblies are identical to those of the metallic core except for the length of reflector pin, shield pin and absorber pins: the lengths are increased because of the taller active core height of the oxide core compared to the metallic core height Material Description of Oxide Core The fuel compositions are obtained from the reference ABR core data at the beginning of equilibrium cycle (BOEC). For the reference ABR core concept, the equilibrium cycle was determined by recycling all discharged TRU. For external makeup feed, TRU recovered from LWR used nuclear fuel is assumed. At the equilibrium cycle, the TRU conversion ratio is ~0.7 and cycle length is days with full power operation. The variation of fuel composition in the irradiated core is simplified. The active core is axially divided into five zones and uniform fuel composition is assumed in each zone. Thus, 15 different fuel compositions are provided in this benchmark: five each for the inner, middle, and outer cores, respectively. A simplified fission products model is also adopted with a pseudo fission product. Natural molybdenum is used to represent the fission products (pseudo fission product). The number densities of coolant and structural materials are identical to those of the metallic core (see Table 21). The fuel pin compositions for the inner, middle, and outer cores at BOEC are provided in Table 29 to Table 31, respectively. All data are obtained for the nominal operating condition by adjusting for the effects of thermal expansion. 34

35 Nuclide Upper boundary from active core bottom (cm) U E E E E E-06 U E E E E E-05 U E E E E E-06 U E E E E E-02 Np E E E E E-05 Pu E E E E E-10 Pu E E E E E-04 Pu E E E E E-03 Pu E E E E E-03 Pu E E E E E-04 Pu E E E E E-04 Am E E E E E-04 Am-242 m E E E E E-06 Am E E E E E-05 Cm E E E E E-06 Cm E E E E E-07 Cm E E E E E-05 Cm E E E E E-05 Cm E E E E E-05 O E E E E E-02 a) Mo E E E E E-04 a) Representative for fission product Table 29 : Number Densities of Inner Core Fuel Pin, Oxide Core, BOEC (atoms/barn-cm) 35

36 Nuclide Upper boundary from active core bottom (cm) U E E E E E-06 U E E E E E-05 U E E E E E-06 U E E E E E-02 Np E E E E E-05 Pu E E E E E-10 Pu E E E E E-04 Pu E E E E E-03 Pu E E E E E-03 Pu E E E E E-04 Pu E E E E E-04 Am E E E E E-04 Am-242 m E E E E E-05 Am E E E E E-04 Cm E E E E E-06 Cm E E E E E-07 Cm E E E E E-05 Cm E E E E E-05 Cm E E E E E-05 O E E E E E-02 a) Mo E E E E E-04 a) Representative for fission product Table 30 : Number Densities of Middle Core Fuel Pin, Oxide Core, BOEC (atoms/barn-cm) 36

37 Nuclide Upper boundary from active core bottom (cm) U E E E E E-06 U E E E E E-05 U E E E E E-06 U E E E E E-02 Np E E E E E-05 Pu E E E E E-10 Pu E E E E E-04 Pu E E E E E-03 Pu E E E E E-03 Pu E E E E E-04 Pu E E E E E-04 Am E E E E E-04 Am-242 m E E E E E-05 Am E E E E E-04 Cm E E E E E-06 Cm E E E E E-07 Cm E E E E E-05 Cm E E E E E-05 Cm E E E E E-05 O E E E E E-02 a) Mo E E E E E-04 a) Representative for fission product Table 31 : Number Densities of Outer Core Fuel Pin, Oxide Core, BOEC (atoms/barn-cm) 37

38 3. Expected Results The following results are expected at the beginning and end of cycle; core multiplication factor (k effective ), sodium void worth, Doppler constant, effective delayed neutron fraction, average nuclide masses or concentrations per each core for end of equilibrium cycle, radial power distribution (integrated over Z for active zone length only), control rod worth (primary and secondary systems) for the following configuration : o all rods inserted (bottom of active zone) The expected heavy metal nuclides are identical to those listed in Table 9 or Table 23. The depletion of the structural materials including boron is ignored in this benchmark. The end of cycle is defined by the core state after one cycle irradiation time, corresponding to: 410 days with full nominal power rating for the large oxide core, 500 days with full nominal power rating for the large carbide core, days with full nominal power rating for ABR metallic and oxide fuel cores, which is equivalent to one-year cycle length with 90% capacity factor. Control rods are supposed to remain at the same position during irradiation. The sodium void worth is defined by the reactivity change between the sodium voided and nominal states such as ρ = ρvoid ρnominal, (1) where the subscripts void and nominal indicate the sodium voided and nominal states, respectively. In this benchmark, the sodium voided state is defined by voiding all sodium in the active core for all defined cores (see active core description in the relevant section). The Doppler constant is defined by ρ ρ high nominal K D =, (2) ln 2 where the subscript high indicates the core state that the fuel temperature in Kelvin is a factor of two of that of the nominal average fuel temperature even if it does not necessarily correspond to a real achievable state for helium bonded fuel (see average fuel temperature in the relevant section). References 1. T. K. Kim, W. S. Yang, C. Grandy and R. N. Hill, Core Design Studies for a 1000 MWth Advanced Burner Reactor, Annals of Nuclear Energy 36 (2009). 38