SAFETY ANALYSIS AND ASSESSMENT FOR TEMPORARY STORAGE OF 106 IRRADIATED VVR-M2 FUEL TYPE ASSEMBLIES IN DALAT REACTOR NGUYEN MINH TUAN AND NGUYEN KIEN CUONG Nuclear Research Institute
Contents 1. Introduction 2. Calculation Model 3. Thermal safety 4. Critical safety 5. Radiological safety 6. Conclusion
1. Introduction (1) ~ 2000 mm Rotating top lid Pool tank VVR-M2 fuel assembly Upper cylindrical shell Extracting well Concrete shielding ~ 6840 mm A Spent fuel storage tank (ex bulk-shielding experimental tank) Core Graphite Thermal column door Door plug Reactor core Reactor pool and shielding structure
1. Introduction (2) Parameter VVR-M2 HEU VVR-M2 LEU Enrichment, % 36 19.75 Average Mass of 235 U in FA, g 40.20 49.70 Fuel Meat Composition U-Al Alloy UO 2 +Al Uranium Density of Fuel Meat, g/cm 3 1.40 2.50 Cladding Material Al alloy (SAV1) Al alloy (SAV1) Fuel Element Thickness (Fuel Meat and 2.50 2.50 Cladding), mm Fuel Meat Thickness, mm 0.70 0.94 Cladding Thickness, mm 0.90 0.78
1. Introduction (3) HEU to LEU conversion project: - Period: 2011 2013 - Number of irradiated fuels: 106 HEU - Average burn-up: 20-22% - SF temporary storage: Reactor pool (3 months) - SF long time storage: SF storage pool (2 years) - LEU fuel loading: New core with 92 FAs - Return of SFs to Russian: 7/2013
1. Introduction (4)
2. Computer codes and Calculation model (1) 69 groups library WIMS-ANL WWR-M2 Decay Library Photon Library Micro cross sections and fission yield Cross sections Library ORIGEN 2.1 Photon spectrum (18 groups) MCNP 5 Gamma dose Material composition, power, t irr Model and flux-to-dose conversion factor
2. Calculation model (2) HEU and LEU fuel in MCNP code
2. Calculation model (3) Reactor hall 3 Concrete wall Reactor hall 2 Reactor hall 1 Concete Reactor Cover Control room Intermediate floor (IF) 1 Reactor Tank Intermediate floor 2 Room No 148 Room 148 Spent Fuel Storage Reactor Core a) b) MCNP Model for Dose rate calculation a) horizontal section, b) vertical section
2. Calculation model (4) MCNP model for LEU core
2. Calculation model (5) MCNP model for fuel temporary storage configuration
2. Calculation model (6) 72 HEU+1IFA+35HEU 45 HEU 106 HEU+1IFA+16LEU 72 HEU + 1IFA
2. Calculation model (6) MCNP Model for long time storage in SP fuel storage pool
3. Decay heat and thermal safety Decay Heat (Watt) of 1 HEU FA, Burn up 30% Cooling time 0 1h 1d 1w 1y 3y 10y Actinide 1.385 0.706 0.422 0.084 0.004 0.003 0.004 FPs 281.065 48.072 9.242 3.142 0.395 0.256 0.191 Total 282.450 48.778 9.663 3.226 0.399 0.260 0.195 Decay Heat (Watt) of 1 LEU FA, Burn up 30% Cooling time 0 1h 1d 1w 1y 3y 10y Actinide 7.051 3.570 2.182 0.503 0.068 0.047 0.049 FPs 270.449 45.991 9.292 3.219 0.489 0.334 0.251 Total 277.500 49.561 11.474 3.722 0.557 0.381 0.299
4. Critical safety Configuration 72 HEU FAs 72 HEU FAs + 35 HEU FAs 45 HEU FAs 106 HEU FAs+ 16 LEU FAs K eff 0.53526 ± 0.015% 0.75278 ± 0.015% 0.78354 ± 0.013% 0.75361± 0.013% Configuration 72 HEU FAs + 1IFA 72 HEU FAs + 35 HEU FAs + 1IFA 34 HEU FAs + 11 LEU FAs 106 HEU FAs+ 16 LEU FAs+ 1IFA K eff 0.53622 ± 0.015% 0.75320 ± 0.015% 0.78916 ± 0.015% 0.75401± 0.013%
5. Radiological safety Activity of HEU fuel assembly (BU-30%) 25000 20000 Activity (Ci) 15000 10000 5000 0 0 5 10 15 20 25 Cooling time (hours)
5. Radiological safety (2) 3.25E+14 3.00E+14 2.75E+14 2.50E+14 Photon Flux (p/cm 2.s) 2.25E+14 2.00E+14 1.75E+14 1.50E+14 Thông lượng 1.25E+14 1.00E+14 7.50E+13 5.00E+13 0 2 4 6 8 1012141618202224262830323436384042444648505254565860626466687072 Cooling time (hours) Photon flux of heavy nuclides and FPs vs. time
5. Radiological safety (3) 3.2E+14 3.0E+14 2.8E+14 2.6E+14 2.4E+14 2.2E+14 Photon Flux (p/cm 2.s) 2.0E+14 1.8E+14 1.6E+14 1.4E+14 1.2E+14 1.0E+14 8.0E+13 6.0E+13 4.0E+13 2.0E+13 1.0HR 24.0HR 7.0D 0.0E+00 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2 2.2 2.4 2.6 2.8 3 3.2 3.4 3.6 Energy (MeV) Photon flux vs. energy
5. Radiological safety (4) Gamma dose rate calculation results (msv/h) (1 day cooling and different water levels) Water Level (cm) Reactor cover Inter. floor 1 Inter. floor 2 Reactor hall 1 Reactor hall 2 Reactor hall 3 Control Room 0 3.860E-06 8.941E-06 5.991E-05 2.101E-07 2.389E-07 2.488E-07 5.679E-08 25 1.130E-05 8.945E-06 6.273E-05 2.120E-07 2.411E-07 2.489E-07 5.685E-08 50 2.776E-05 9.051E-06 6.379E-05 2.172E-07 2.528E-07 2.512E-07 5.814E-08 75 7.976E-05 9.753E-06 6.379E-05 2.352E-07 2.728E-07 2.741E-07 6.329E-08 100 2.754E-04 9.833E-06 6.399E-05 2.983E-07 3.576E-07 3.617E-07 6.765E-08 125 7.702E-04 9.931E-06 6.525E-05 4.545E-07 5.538E-07 5.572E-07 7.406E-08 150 2.419E-03 1.009E-05 6.581E-05 1.051E-06 1.293E-06 1.253E-06 1.143E-07 175 7.238E-03 1.676E-05 6.727E-05 2.769E-06 3.516E-06 3.402E-06 2.295E-07 200 2.421E-02 2.727E-05 8.349E-05 8.831E-06 1.081E-05 1.069E-05 6.191E-07 225 7.660E-02 7.315E-05 1.249E-04 2.856E-05 3.608E-05 3.558E-05 1.897E-06 250 2.584E-01 2.435E-04 3.356E-04 1.056E-04 1.308E-04 1.270E-04 7.942E-06 275 8.757E-01 9.030E-04 1.254E-03 3.873E-04 4.780E-04 4.642E-04 2.975E-05 300 2.955E+00 3.440E-03 5.610E-03 1.460E-03 1.800E-03 1.730E-03 1.000E-04 325 1.492E+01 1.936E-02 3.438E-02 8.110E-03 9.840E-03 9.380E-03 5.700E-04 385 2.411E+01 3.176E-02 5.803E-02 1.325E-02 1.606E-02 1.531E-02 9.500E-04 465 2.485E+01 3.211E-02 5.765E-02 1.342E-02 1.630E-02 1.557E-02 9.700E-04 625 Total 106 HEU 92 LEU 3.699E+01 2.402E+01 1.297E+01 4.408E-02 3.137E-02 1.271E-02 7.326E-02 5.706E-02 1.620E-02 1.873E-02 1.313E-02 5.610E-03 2.295E-02 1.587E-02 7.070E-03 2.198E-02 1.516E-02 6.830E-03 1.350E-03 9.500E-04 4.000E-04
5. Radiological safety (5) Gamma dose rate calculation results (msv/h) (7 day cooling and different water levels) Water levels (cm) Reactor cover Inter. floor 1 Inter. floor 2 Reactor hall 1 Reactor hall 2 Reactor hall 3 Control Room 0 3.191E-06 6.179E-06 5.030E-05 1.652E-07 1.873E-07 1.913E-07 3.589E-08 50 2.145E-05 6.280E-06 5.037E-05 1.711E-07 1.969E-07 1.974E-07 4.146E-08 100 1.941E-04 6.368E-06 5.080E-05 2.246E-07 2.658E-07 2.680E-07 4.854E-08 150 1.756E-03 8.121E-06 5.153E-05 7.464E-07 9.408E-07 9.240E-07 9.340E-08 200 1.711E-02 1.926E-05 5.917E-05 5.895E-06 7.438E-06 7.329E-06 4.157E-07 250 1.732E-01 1.470E-04 1.972E-04 6.364E-05 8.047E-05 7.793E-05 4.594E-06 325 5.538E+00 5.880E-03 9.390E-03 2.440E-03 3.030E-03 2.920E-03 1.800E-04 385 1.329E+01 1.442E-02 2.386E-02 5.910E-03 7.310E-03 7.050E-03 4.300E-04 465 1.375E+01 1.467E-02 2.387E-02 6.030E-03 7.470E-03 7.220E-03 4.400E-04 625 Total 106HEU 92LEU 2.033E+01 1.325E+01 7.075E+00 2.013E-02 1.431E-02 5.820E-03 2.903E-02 2.314E-02 5.890E-03 8.450E-03 5.860E-03 2.590E-03 1.056E-02 7.240E-03 3.320E-03 1.021E-02 6.990E-03 3.220E-03 6.100E-04 4.300E-04 1.800E-04
5. Radiological safety (6) Gamma dose rate calculation results (msv/h) (1 hour cooling and completely loosing water) Reactor cover Inter. floor 1 Inter. floor 2 Reactor hall 1 Reactor hall 2 Reactor hall 3 Control Room 106 HEU FAs 361.272 0.52409 1.52921 0.12914 0.16290 0.15902 0.01158 New core with 92 LEU FAs 142.366 0.10460 0.09495 0.04637 0.05998 0.05864 0.00319 Total 503.637 0.62868 1.62416 0.17551 0.22288 0.21767 0.01477
6. Conclusions - When the water level of reactor pool is maintained not loosing to exceed 100cm, the gamma dose caused by HEU FAs does not effect employees working in reactor hall. - In principle, the longer cooling time the higher of safety, to avoid risk to employees fuel unloading for core conversion should be conducted after shutting down reactor 2 weeks. - Critical calculations show that with the configurations when storing HEU spent FAs, the risk to get criticality is impossible to happen.