GENES4/ANP2003, Sep. 15-19, 2003, Kyoto, JAPAN Paper 1018

Full MOX Core Design in ABWR

Toshiteru Ihara1*, Takaaki Mochida2, Sadayuki Izutsu3 and Shingo Fujimaki3 1Nuclear Power Department, Electric Power Development Co., Ltd., Tokyo, 104-8165, Japan 2Nuclear Plant Engineering Department, Hitachi, Ltd., Hitachi, Ibaraki, 319-1188, Japan 3Core Design Group, Global Nuclear Fuel Japan Co., Ltd., Yokosuka, Kanagawa, 239-0836, Japan

Electric Power Development Co., Ltd. (EPDC) has been investigating an ABWR plant for construction at Oma-machi in Aomori Prefecture. The reactor, termed FULL MOX-ABWR will have its reactor core eventually loaded entirely with mixed-oxide (MOX) fuel. Extended use of MOX fuel in the plant is expected to play important roles in the country’s nuclear fuel recycling policy. MOX fuel bundles will initially be loaded only to less than one-third of the reactor, but will be increased to cover its entire core eventually. The number of MOX fuel bundles in the core thus varies anywhere from 0 to 264 for the initial cycle and, 0 to 872 for equilibrium cycles. The safety design of the FULL MOX-ABWR briefly stated next considers any

probable MOX loading combinations out of such MOX bundle usage scheme, starting from full UO2 to full MOX cores.

KEYWORDS: Full MOX, ABWR, Core Design

I. Introduction Table 2 respectively.

The core design from full UO2 to full MOX loaded of (1) The MOX bundle uses the well-proven design of STEP-2 1) 2) 3) ABWR has been performed. The MOX fuel is 8x8 UO2 bundle (50GWd/t maximum exposure) having much bundle configuration with a large central water rod, with operational experience. The bundle features a large water 40GWd/t maximum burnup, and it is compatible with 9x9 rod in the center of 8x8 fuel rod configuration, as shown high burnup UO2 fuel. The shutdown and thermal margins of in Fig.1. the MOX core, from partially- to fully- loaded, are (2) The MOX bundle maximum exposures conservatively comparable to that of full UO2 core. Safety analyses based are selected to be 40 GWd/t based on the MOX on MOX loaded core characteristics and MOX fuel property irradiation experience. This maximum exposure is equal have shown its conformity to the design criteria in Japan. to that of STEP-I UO2 bundle. This paper shows the safety design for full MOX cores of (3) The MOX fuel bundle consists of MOX fuel rods using

ABWR. UO2-PuO2 and UO2 fuel rods using UO2-Gd2O3 as a fuel material. The bundle average fissile material content is II. Full MOX Core selected to be about 2.9wt% of fissile plutonium (Puf) 1. Design Concept content and about 1.2wt% of U-235 enrichment for the One of the ABWR core features is a wider fuel bundle conditions of 13-month cycle length and the reference pitch. Specifications of fuel bundle are the same as those of plutonium composition (67 wt% Puf ratio). the current BWR lattice. As a result, the non-boiling water area outside the channel box (bypass flow area) increases in the ABWR core. The bypass flow area can thermalize Table 1 Basic specification of core design neutrons more effectively before absorption by the fuel Items Specification material. The wider fuel pitch of the ABWR core decreases Core the absolute value of void reactivity coefficient and Type Advanced Boiling Water increases the shutdown margin, which collectively makes Reactor ABWR well-adopted for loading its core fully with MOX (ABWR) fuel. Thermal power (MW) 3,926 The main specifications are not changed from those of the Core flow (t/h) 52.2x103 (100%rated) standard ABWR, such as 3926MWt thermal power, 872 fuel Core pressure (MPa[abs]) 7.17 bundles and 205 control rods. The main design concepts of Number of fuel bundles 872 MOX fuel are the following, and the basic specifications of Number of control rods 205 core are shown in Table 1 and those of fuel are shown in Fuel bundle pitch (cm) 15.5

* Corresponding author, Tel. +81-3-3546-2211, Fax. +81-3-3546- 2805, E-mail: [email protected] Table 2 Basic specification of fuel design lifetime. Also, though MOX fuel tends to become higher rod internal pressure owing to increased FP gas and He gas Items Specification release, the rod internal pressure remains still equivalent to Fuel assembly 9x9 UO fuel rod at the end of lifetime due to increased gas Lattice configuration 2 plenum volume for MOX rod. The rod internal pressure of MOX fuel 8x8 the MOX fuel is shown in comparison with 9x9 UO2 fuel in UO2 fuel (STEP-III 9x9) 9x9 Average 235U content *1(wt%) Fig.2. MOX fuel 1.2 The neutron multiplication factor (k-infinity) of MOX fuel is shown in Fig.3 in comparison with 9x9 UO2 fuel. UO2 fuel (STEP-III 9x9) 3.8 Average Puf content *1(wt%) The change in k-infinity with exposure is smaller for MOX MOX fuel 2.9 fuel than that for UO2 fuel. This, in turn, leads to a smaller

UO2 fuel (STEP-III 9x9) - bundle peaking of full MOX core than full UO2 core and 1/3 Maximum exposure (MWd/t) MOX core (i.e. consist of 512 9x9 UO2 fuels and 360 MOX MOX fuel 40,000 fuels). This brings about the decrease of the maximum

UO2 fuel (STEP-III 9x9) 55,000 control rod worth and the mitigation of the shutdown margin Batch averaged exposure (MWd/t) decrease for full MOX core. MOX fuel 33,000 Figure 4 shows the fuel loading pattern of full MOX UO2 fuel (STEP-III 9x9) 45,000 equilibrium core. Shutdown margin, Maximum Linear Heat Number of fuel rods Generation Rate (MLHGR), and Minimum Critical Power MOX fuel 60 Ratio (MCPR) for the equilibrium cycle are shown in Fig.5 UO fuel (STEP-III 9x9) 74 *3 2 the comparison among full UO2 core, 1/3 MOX core and full Pellet material MOX core. MOX fuel UO2-PuO2(MOX rods) The shutdown and thermal margins of the MOX core, UO2-Gd2O3(UO2 rods) from partially- to fully- loaded, are comparable to that of full UO2 fuel (STEP-III 9x9) UO2,UO2-Gd2O3 UO2 core. These results satisfy the design targets and the Cladding outside diameter (mm) operational limits. 12.3 MOX fuel 11.2 UO2 fuel (STEP-III 9x9) Control rod Cladding wall thickness *2(mm) MOX fuel 0.86 43222234 UO2 fuel (STEP-III 9x9) 0.71 Cladding Material 31113 MOX fuel Zirc-2(with Zr liner) 211 1112 UO2 fuel (STEP-III 9x9) Zirc-2(with Zr liner) Number of water rods 2 112 MOX fuel 1 w 2 12 UO2 fuel (STEP-III 9x9) 2 1 Water rod outside diameter (mm) 2111 112 34.0 MOX fuel 311 13 UO2 fuel (STEP-III 9x9) 24.9 43222234 *1 For reload fuel W ： Water rod *2 Including Zirconium liner thickness of about 0.1mm 1 ：4 MOX fuel rod *3 Including 8 partial length rods (Pu content 4<3<2<1) ： UO2 fuel rod containing Gd2O3 2. Fuel and Core Design Rod arrangement of the MOX fuel bundle is shown in Fig.1 Rod arrangement of MOX fuel bundle Fig.1. Of the 60 fuel rods, 48 contain MOX and the rest UO2 bearing Gd2O3. Four kinds of MOX rods with different plutonium contents in the bundle are employed in order to reduce the local peaking. Although MOX fuel tends to become higher pellet temperature owing to lower thermal conductivity and increased FP gas release compared with

UO2 fuel, the pellet center temperature remains sufficiently lower against the fuel melting temperature through the 6 8.0 ) 80 k Full MOX fuel core Coolant pressure

△ 5 9X9 UO2 fuel core % ( 4 1/3 MOX fuel core in

6.0 g 60 3 2 a) 2 Design target 4.0 40 1

(kg/cm 0 Shutdown mar Shutdown 01234567891011 Cycle exposure (GWd/t) 2.0 MOX fuel 20 9X9 UO2 fuel (a) Shutdown margin Fuel rodFuel internal pressure (MPa[abs])

0.0 0 0 20406080 50 Full MOX fuel core Peak pellet exposure (GWd/t) Operating limit 45 9X9 UO2 fuel core Fig.2 Fuel rod internal pressure 1/3 MOX fuel core 40 1.3 35 1.2 Hot operation 30 MLHGR (kW/m) 25 1.1 01234567891011 1.0 Cycle exposure (GWd/t) (b) MLHGR 0.9 MOX fuel bundle 0.8 9X9 UO2 fuel bundle 1.9 Full MOX fuel core 1.8 Neutron multiplication factor multiplication Neutron 0.7 9X9 UO2 fuel core 1.7 1/3 MOX fuel core 0 1020304050 1.6 Exposure (GWd/t) 1.5

Fig.3 Neutron multiplication factor MCPR 1.4 Operating limit Full MOX 1.3 1.2 1/3 MOX 1.1 01234567891011 Cycle exposure (GWd/t) (c) MCPR

Fig.5 Comparison of core performance in equilibrium

3. Characteristics of MOX core 3.1 Reactivity coefficient and dynamic parameter The MOX core is characterized by an increase in the absolute void coefficient due to large neutron absorption cross section of plutonium relative to uranium. Void MOX fuel coefficient, Doppler coefficient, delayed neutron fraction Fresh reload bundle and prompt neutron lifetime change depending on the MOX Bundle in 2nd cycle of operation fuel loading fraction (Fig.6). While the void coefficient of Bundle in 3rd cycle of operation the full MOX core is about 20% larger in absolute value Bundle in 4th cycle of operation than the full UO2 core, the delayed neutron fraction of the Fig.4 Fuel loading pattern of full MOX equilibrium cycle full MOX core is about 20% smaller than the full UO2 core. The absolute values of Doppler coefficient of MOX loaded neutron absorption cross section, in the plutonium isotopes core is equivalent to that of 9x9 UO2 core. becomes small as Puf ratio becomes large. The MOX bundle can be used regardless of its plutonium 2 Void reactivity coefficient vector to adjust the plutonium content of MOX bundle according to the reactivity compensation design. Shutdown Doppler reactivity coefficient margin, MLHGR, and MCPR of full MOX core with Delayed neutron fraction reference plutonium vector in equilibrium cycle are shown 1 in Fig.9 compared with other vector (Puf ratio: 75 and 62 wt%). These results satisfy the design target or operational limit. Relative value

0 1.20 0 1/3 3/3 MOX fule (Puf ratio 67wt%) MOX fule (Puf ratio 77wt%) (9X9 UO2 core) (1/3 MOX core) (Full MOX core) 1.15 MOX fule (Puf ratio 58wt%) UO2 fuel (batch averaged exposure 33GWd/t) MOX fuel loading fraction 1.10

Fig.6 Safety parameter vs. MOX loading fraction 1.05 Hot operation 1.00 3.2 Influence of plutonium isotopic composition 4) k-infinity 0.95 (1) Reactivity compensation design of MOX fuel 22GWd/t Since the plutonium mixed with the uranium come from 0.90 (EOC) the reprocessing spent fuels, the isotope composition of 0.85 plutonium (plutonium vector) varies depending on initial 0 10203040 enrichment, burn-up, cooling period, etc. of the reprocessed exposure (GWd/t) UO2 fuels. The main isotopes of plutonium obtained by the reprocess are Pu-239 and Pu-241 which are fissile, Pu-238, Fig.7 Neutron multiplication factor vs. exposure Pu-240 and Pu-242 which are fertile. The plutonium vector slightly changes also after reprocessing, since Pu-241 among these isotopes decays in short period (half life of about 14.4 years) and becomes Am-241 of a non-fissile. 8 As the uranium of MOX base matrix is the tail uranium of 7 Pu content (matrix U-235:0.2wt%) enrichment to achieve effective use of plutonium, U-235 content of MOX pellet has the range from about 0.2wt% to 6 Puf content (matrix U-235:0.2wt%) about 0.4wt% depending on the enrichment process of uranium. In the fabrication of MOX fuel, the plutonium 5 content of MOX is adjusted depending on the plutonium 4 vector, so that the MOX bundle is able to achieve same design discharge exposure. This is called the reactivity 3 compensation design. The following are considered in the 2 reactivity compensation design. (a) The plutonium average content of MOX bundle is 1 adjusted to have the target k-infinity at 22 GWd/t 0 corresponding to the core average exposure at the end of Bandle average Pu and Puf content (wt%) cycle (EOC). The target k-infinity is equal to the k- 55 60 65 70 75 80 85 infinity of UO2 bundle at 22 GWd/t. (b) All MOX rods have uniform axial plutonium contents to Puf ratio (wt%) simplify the MOX fuel fabrication. A design policy of Fig.8 Pu and Puf content vs. Puf ratio having only four different plutonium contents is maintained irrespective of a change in plutonium vector from reference. Figure 7 shows the k-infinity changes with exposure depending on the plutonium vector of MOX. Figure 8 shows the average plutonium and Puf content of MOX bundle with the Puf ratio (range is about 58-77wt%). The plutonium content decreases as the Puf ratio becomes large. This reason is that the ratio of Pu-240, that has large plutonium vector. The influence of plutonium vector on 5 Puf ratio 67wt% delayed neutron fraction is shown in Fig.11. Puf ratio 75wt% k) Although the influence of plutonium vector on dynamic 4 Puf ratio 62wt% Δ coefficient is relatively small as shown in Fig.10, the 3 influence is taken into account in the safety evaluation. In 2 the safety analysis of full MOX core, the dynamic void coefficient in the reference plutonium vector is multiplied by design target ≧1% Δk 1 a factor 1.04, and dynamic Doppler coefficient by 0.96. As apparent from Fig.10, these factors are conservatively 0 Shutdown margin (% determined. -1 The influence of plutonium vector on the scram reactivity 01234567891011 worth is shown in Fig.12. The difference of static control rod worth (in %∆k unit) between MOX core and UO core is cycle exposure (GWd/t) 2 small because the control rod is inserted in a water gap away (a) Shutdown margin from the fuel bundle in BWR. The dynamic control rod worth (in dollar unit) of MOX core is larger than that of UO 50 2 operation limit 44.0kW/m core because the delayed neutron fraction of MOX core is smaller than that of UO core. The design scram curve has 40 2 enough margin in either case. 30 ABWR,EOC, core average void fraction 40% Puf ratio 67wt% 20 Puf ratio 77wt% 10 Puf ratio 58wt% MLHGR (kW/m) 10 5 0 01234567891011 0 cycle exposure (GWd/t)

(b) MLHGR more nagative ← neutron fraction) (%) -5 -4% 1.9 Puf ratio 67wt% 1.8 Puf ratio 75wt% -10 Puf ratio 62wt% 1.7 Relative value of (void coefficient / delayed 55 60 65 70 75 80 85 1.6 Puf ratio (wt%) 1.5

MCPR Fig.10(a) Dynamic void coefficient vs. Puf ratio 1.4 operation limit 1.32(from EOC-2GWd/t to EOC) 1.3 1.25(other period) ABWR,EOC,cold 1.2 10 1.1 01234567891011 +4% 5 cycle exposure (GWd/t) (c) MCPR 0 Fig.9 Comparison of equilibrium core performance

-5 more nagative (2) Influence on reactivity coefficients ←

The influence of plutonium vector on dynamic void delayed neutron fraction) (%) coefficient and dynamic Doppler coefficient (coefficient

Relative value of (Doppler coefficient / -10 divided by delayed neutron fraction) are shown in Fig.10(a) and Fig.10(b) respectively. In Fig.10, the Y-axis indicates 55 60 65 70 75 80 85 relative difference in dynamic coefficient expressed by Puf ratio (wt%) (A-B)/B where A is a dynamic coefficient for plutonium Fig.10(b) Dynamic Doppler coefficient vs. Puf ratio vector X and B is the same coefficient for the reference account of power distributions for cores covering both full

ABWR,EOC, 9x9 UO2 core and full MOX core, and additionally, those core average void fraction 40% falling between these cores. The largest negative void 10 coefficient, observed throughout the lifetime of the core for each for full 9x9 UO2 core and full MOX core, is selected. Table 3.shows the decay ratios of core stability and 5 regional stability at the condition of minimum pump speed and maximum reactor power. Those are calculated by the frequency domain code. Although the decay ratios of full 0 MOX core are larger than those of 9x9 UO2 core, these are (%) enough smaller than criterion 1.0.

-5 Table 3 Decay ratio for core stability and regional stability Core stability Regional stability Full MOX core 0.68 0.55 -10 UO2 core 0.37 0.26 Relative value of delayed neutron fraction 55 60 65 70 75 80 85 2. Abnormal transient during plant operation Puf ratio (wt%) One of the features of MOX loaded core is that the Fig.11 Delayed neutron fraction vs. Puf ratio absolute value of void coefficient become larger according ABWR,EOC to the increase of MOX loading fraction. The larger void -50 coefficient in absolute value brings the rapider decrease of Puf ratio 67wt% Puf ratio 75wt% reactor power in the case that the reactor power rises Puf ratio 62wt% following the void fraction in the core increases by some -40 UO2 core causes. The MOX fuel loading brings higher self-control design curve ability of reactor power. In the event in which void fraction increases such as -30 “partial loss of coolant flow”, the large void coefficient in absolute value accelerates the decrease of power and mitigates the increase of ∆MCPR (index of thermal margin). -20 On the other hand, in the event in which void fraction decreases such as “generator load rejection without turbine

Scram reactivity (dollar unit) bypass”, the large void coefficient in absolute value -10 accelerates the increase of power and ∆MCPR. The ∆MCPR in “loss of feed water heating” and “generator load rejection without turbine bypass” that are 0 void fraction decreasing event and the severest event in MCPR are shown in Fig.13. 0.0 0.2 0.4 0.6 0.8 1.0 “Loss of feed water heating” is the event in which reactor Control rod insert fraction power gradually rises according to the slow decrease of void Fig.12 Scram reactivity worth vs. Puf ratio fraction in the core due to the slow decrease of inlet coolant temperature. Since the reactor power slowly rises and the III. Safety evaluation maximum power is limited by the scram of “neutron flux 1. Stability high signal”, the ∆MCPR hardly depends on the MOX The thermal-hydraulic characteristic of MOX bundle is loading fraction, that is, the magnitude of void coefficient comparable to that of 9x9 UO bundle. The core loaded with 2 and the difference between scram curves for early stage and both kinds of bundle has intermediate void coefficient end stage in cycle shown in Fig.14. between those of the core loaded with single kind of bundle “Generator load rejection without turbine bypass” is the respectively. Therefore the core stability and regional event in which reactor power rapidly rises according to the stability of the core loaded with both kinds of bundle is fast decrease of void fraction by the turbine control valve intermediate between the cores loaded with single kind of fast closure. The reactor power of MOX loaded core rapidly bundle. rises compared with 9x9 UO loaded core because of large The core stability and regional stability about full MOX 2 negative void coefficient. The ∆MCPR depends on the loaded core and full 9x9 UO loaded core are 2 MOX loading fraction, that is, the magnitude of void representatively evaluated assuming the conservative power coefficient and the difference between scram curves for distribution and void coefficient. The power distribution is early stage and end stage in cycle shown in Fig.14. determined to provide conservative assumptions taking In the core applied to the scram curve of early stage in cycle, the ∆MCPR of “generator load rejection without 3. Accident analysis turbine bypass” is smaller than that of “Loss of feed water “Loss of the coolant accident (LOCA)”,“All pump trip heating” regardless of MOX loading fraction, therefore the accident (APTA)” and “Main steam line break accident ∆MCPR determining event is “Loss of feed water heating”. (MSLBA)” are the events in where the void fraction In the core applied to the scram curve of end stage of cycle, increases. The full MOX core with negative void coefficient the ∆MCPR determining event changes from “Loss of feed larger than that of full 9x9 UO2 core tends to reduce the water heating” to “generator load rejection without turbine reactor power further during event. MOX fuel has the bypass” at about 1/3 MOX loading fraction (360 MOX fuel tendency to make the fuel cladding temperature higher, loading). because the gap conductance and the pellet thermal

The different two operational limit MCPR (OLMCPR) conductivity are small compared with the UO2 fuel. For are selected to secure adequate operational thermal margin. MOX loaded core, the time dependent reactor power

One is used for the core applied to the scram curve of early identical to the UO2 core is adopted conservatively. The fuel stage in cycle regardless of MOX loading fraction, and the cladding temperature of LOCA is evaluated for “High core applied to the scram curve of end stage in cycle below Pressure Core Flooder system (HPCF) pipe break” as the 360 MOX fuel loading. Another is used for the core applied severest case for the drop of reactor water level. The to the scram curve of end stage in cycle above 360 MOX cladding surface temperature rises due to boiling transition, fuel loading. but the core is always covered with two phase coolant The void coefficients of 360 MOX loaded core and of full mixture over the whole transient period. The peak cladding MOX loaded core in equilibrium cycle end are used in the temperature (PCT) is about 600oC, satisfying the criterion of ∆MCPR evaluation of the core below 360 MOX loading and 1200oC, although PCT of MOX core becomes a little higher the core above 360 MOX loading respectively as the void than UO2 core. reactivity feedback is conservatively estimated. 2% for 360 The fuel enthalpy response of “Control Rod Drop MOX loading core and 4% for full MOX loaded core are Accident (CRDA)” is shown in Fig.15 as a representative added to the void coefficient in the reference vector to event of transient and accident for control rod system, where appropriately consider the influence of plutonium vector. the dropping control rod worth is assumed to be 1.3% ∆k. The maximum fuel enthalpy in the full MOX core is almost 0.4 equivalent to that of the UO2 core, because Doppler Generator load rejection (BOC) reactivity effect in dollar unit of full MOX core is more 0.3 Generator load rejection (EOC) negative. Loss of feed water heating Full MOX fuel core 0.2 9x9 UO fuel core MCPR 2 △ 0.1

0.0 0 0.2 0.4 0.6 0.8 1 MOX loading fraction

Fig.13 ∆MCPR vs. MOX loading fraction

-50 Fuel enthalpy

For early stage in cycle -40 For end stage in cycle

-30 Time (sec)

-20 Fig.15 Fuel enthalpy response on CRDA

Scram reactivity (dollar unit) IV. Others -10 An increase of MOX fuel loading fraction causes the decrease in the boric acid water reactivity worth of the 0 standby liquid control (SLC) system. This decrease has been 0.0 0.2 0.4 0.6 0.8 1.0 compensated by the increase in the SLC tank capacity from Control rod insert fraction 29m3 to 36m3. Fig.14 Design scram curve Since the control rod is inserted in a water gap, which is full MOX core as well as in the case of UO2 core. The out of the fuel bundle in BWR, the influence of MOX fuel parameters used for safety analysis are selected loading on the control rod worth is not so large. This appropriately considering MOX fuel and MOX loaded core influence become smaller in the ABWR because the ABWR characteristics. As the result of such safety analysis, it is has a wider bundle pitch than BWRs. The total rods worth of confirmed that criteria concerning to safety evaluation are a current designed full MOX core is almost comparable to satisfied in the case of MOX loaded to full MOX core as that of full UO2 core in the ABWR as shown in Fig.16. well as in the case of UO2 core. To provide for the decrease of reactivity due to decay of Pu-241 associated with unpredictable delays of loading, the Nomenclature maximum core flow rate in full power operation has been Puf ratio: increased from 111% in the existing ABWRs to 120% in the (Pu-239+Pu241)/(total Pu+Am-241)*100wt% FULL MOX-ABWR. Pu content: (total Pu+Am-241)/(total Pu+Am-241+total U)*100wt% cold Puf content: 20 (Pu-239+Pu241)/(total Pu+Am-241+total U)*100wt%

k) : total control rod reactivity worth

Δ : maximum control rod reactivity worth 15 References 1) Y.Kinoshita, et al., “Design of Full MOX Core in ABWR”, Karyoku-Genshiryoku-Hatsuden (J. Thermal 10 and Nuclear Power Engineering Soc.), Vol.50 No.2, 62 (1999) [in Japanese] 2) S.Izutsu, et al., “PROGRESS OF FULL MOX CORE 5 DESIGN IN ABWR”, International Symposium on MOX Fuel Cycle Technologies, IAEA-SM-358/27, Control rod reactivity worth (% 0 Vienna (1999)

9x9 UO2 core 1/3MOX core Full MOX core 3) S.Izutsu, “Core design of Full MOX-ABWR”, 32-th Seminar of reactor physics in summer, Yufuin, Japan, Fig.16 Total control rods reactivity worth July 31- August 2, 2000, p14-26 (2000) [in Japanese] 4) Hitachi Ltd., “BWR About Pu isotope change of V. C o nc l usi o n MOX loaded core in the Full MOX-ABWR”, HLR-067 (1999) [in Japanese] As the result of thermal-mechanical MOX fuel rod analysis and MOX loaded core analysis, it is confirmed that design criteria are satisfied in the case of MOX loaded to