Model updates and performance evaluations on fuel performance code FEMAXI-8 for light water reactor fuel analysis

Figures & data

Figure 1. Configuration of finite elements and nodes representing an axial node of fuel stack.

Figure 2. Comparison of mechanical behaviors of the fuel rod irradiated at the DR3 reactor in RISO program [5] simulated with two different relocation model sets FEM7 and FEM8: (a) linear heat rate and P/C contact pressure and (b) fuel axial stresses in the 1st and 2nd finite elements.

Table 1. Summary of irradiation test conditions treated in the validation of FEMAXI-8 [4]

	Fuel type^c
	UO2	MOX	Gd-doped
Burnup [GWd/tU]	3–99	15–72	25–95
Max. LHR^a [W/cm]	146–585	201–469	174–392
Rod pressure^b [MPa]	0.1–3.4	0.1–2.6	1.0–2.6
He ratio in rod filler gas^b	0.0–1.0	1.0	0.5–1.0
P/C diametral gap^b [μm]	42–508	75–290	49–186

^aRod average linear heat rate that a rod experienced during its irradiation period.

^bInitial condition.

^cCladding material is stress relief annealed Zircaloy-4 or recrystallization annealed Zircaloy-2 in most cases.

Table 2. Groups of validation-calculation cases [4]

Validation case group ID	Measurement data focused in validation
PCT/BOL	PCT^a at the BOL
PCT/LowBU	PCT at low burnup <5 GWd/tU
PCT/UO2, PCT/MOX, PCT/Gd	PCT of UO2, MOX, and Gd-doped rods for a whole irradiation period
FGR/UO2/Base, FGR/MOX/Base, FGR/Gd/BASE	FGR from UO2, MOX, and Gd-doped rods
FGR/UO2/Ramp, FGR/MOX/Ramp	FGR from UO2 and MOX rods subjected to a power ramp
FreeVol	Rod free volume at the EOL
HoopStrain	Hoop strain increment in cladding tube induced by a power ramp
Oxidation	Cladding oxide thickness at the EOL

^aPellet center temperature.

Figure 3. Comparison of fuel center temperatures between calculation results and measurements: (a) ‘PCT/BOL’ group, (b) 0–0.2 GWd/tU of ‘PCT/LowBU’ group, (c) 0.2–5.0 GWd/tU of ‘PCT/LowBU’ group, (d) ‘PCT/UO2’ group, (e) ‘PCT/MOX’ group, and (f) ‘PCT/Gd’ group.

Figure 4. Comparison of fission gas releases between calculation results and measurements. In the case of a ramp irradiation test conducted after a base irradiation on the tested fuel rod, fission gas release during the base irradiation is excluded and solely the incremental fission gas release during the ramp irradiation is referred to.

Figure 5. Comparison of (a) rod free volume after base-irradiation and (b) cladding deformation induced by a power ramp between calculation results and measurements: (a) ‘FreeVol’ group and (b) ‘HoopStrain’ group, respectively.

Figure 6. Comparison of cladding oxide thickness after base irradiation between calculation results and measurements for ‘oxidation’ group. Data points are identified with test IDs defined in the NSRR experiment program.

Figure 7. Benchmark result on computation speeds of FEMAXI-8 and FEMAXI-7 with 118 calculation cases out of the irradiation test cases treated in the validation of FEMAXI-8 Y.

Bagger C, Carlson H, Knudson P. Details of design irradiation and fission gas release for the Danish UO2-Zr irradiation test 022. Denmark: Risø National Laboratory; 1978. RISO-M-2152.

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Udagawa Y, Yamauchi A, Kitano K, et al. Development of fuel performance code FEMAXI-8—model improvements for light water reactor fuel analysis and systematic validation. Japan: Japan Atomic Energy Agency; 2018. JAEA- Data/Code 2018-016. in Japanese.

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