Derivation of the source term, dose results and associated radiological consequences for the Greek Research Reactor – 1

Marine and atmospheric transport modeling supporting nuclear preparedness in Norway: Recent achievements and remaining challenges

Numerical transport models are important tools for nuclear emergency decision makers in that they rapidly provide early predictions of dispersion of released radionuclides, which is key information to determine adequate emergency protective measures. They can also help us understand and describe environmental processes and can give a comprehensive assessment of transport and transfer of radionuclides in the environment. Transport of radionuclides in air and ocean is affected by a number of different physico-chemical processes. Along with uncertainty arising from the input data, the model estimates will therefore involve a combination of numerous uncertain factors, caused by knowledge gaps and assumptions in the model system. As discussed in this paper, the major sources to uncertainty affecting the model results are release descriptions, driving data, process descriptions and parameters. Here, we give a synthesis of the most important improvements in atmospheric and marine models achieved through the CERAD programme. In the atmospheric transport model, an important improvement has been inclusion of uncertainties in the dispersion estimates. Recent developments also include adaption to high resolution forcing data and ensemble forecasts, inversion methods and long term analyses. Case studies clearly show improved predictions from ensemble mean values compared to single deterministic runs, and promises for future upgrades of preparedness decision support systems. A major improvement in the marine model system was implementation of dynamic speciation including transformation of species, identifying particle size and parameterizations to be key factors affecting radionuclide distribution. The model system was further developed in a case study involving the impact of changing environmental factors on the transport of aluminium river run-off to an estuary in southeastern Norway. Suggestions for future improvements include implementation of an operational preparedness model for marine transport, better quantification of uncertainties using ensemble methods and improved source identification with further development of inverse transport.

Coupling of ASTEC V2.1 and RASCAL 4.3 Codes to Evaluate the Source Term and the Radiological Consequences of a Loss-of-Cooling Accident at a Spent Fuel Pool

This paper deals with a general methodology to evaluate the Source Term (ST) and the Radiological Consequences (RC) of a hypothetical Severe Accident at a Fukushima-like Spent Fuel Pool by coupling ASTEC 2.1 and RASCAL 4.3 severe accident and consequence projections codes, respectively. The methodology consists of two steps: the ST provided by a priori simulation performed by ASTEC V2.1 code was used as input to RASCAL 4.3 code to make an RC analysis. This approach was developed as a preparatory study for the Management and Uncertainties in Severe Accident (MUSA) H2020 European Project, coordinated by CIEMAT, where the ENEA's Nuclear safety laboratory is committed to perform an analysis on a Fukushima-like SFP with the aim to apply innovative management of SFP accidents (WP6) to mitigate the RC of the accident itself. To perform the RC studies that could have an impact on Italy, a Fukushima-like SFP was assumed located in one of the Italian cross-border NPPs. The weather data adopted are both standard and real hourly meteorological data taken from more than one geographical location. The results of the RC for 96 hours of ST release in a range of 160 km from the emission point are reported in terms of Total Effective Dose Equivalent, Thyroid dose, and Cs-137 total ground deposition. The mitigating effect on ST and on RC of the cooling spray system (CSS) actuated with several pH values (i.e., 4,7,10) was also investigated.

Source Term Derivation and Radioactive Release Evaluation for JRTR Research Reactor under Severe Accident

The source term for the JRTR research reactor is derived under an assumed hypothetical severe accident resulting in generation of the most severe consequences. The reactor core is modeled based on the reactor technical design specifications, and the fission products inventory is calculated by using the SCALE/TRITON depletion sequence to perform burnup and decay analyses via coupling the NEWT 2-D transport lattice code to the ORIGEN-S fuel depletion code. Fifty radioisotopes contributed to the evaluation, resulting in a source term of 3.7 × 1014 Bq. Atmospheric dispersion was evaluated using the Gaussian plume model via the HOTSPOT code. The plume centerline total effective dose (TED) was found to exceed the IAEA limits for occupational exposure of 0.02 Sv; the results showed that the maximum dose is 200 Sv within 200 m from the reactor, under all the weather stability classes, after which it starts to decrease with distance, reaching 0.1 Sv at 1 km from the reactor. The radiation dose plume centerlines continue to the exceed international basic safety standards annual limit of 1 mSv for public exposure, up to 80 km from the reactor.

Determination of radiological source term of CHASHMA-1 NPP during LOCA

The CHASHMA Nuclear Power Plant unit 1 is known as CHASNUPP-1. The CHASNUPP-unit 1 is a 996 MWth intermediate type pressurised water reactor that began commercial operation in June 2000 in Pakistan. The CHASNUPP unit 1 is a conventional two loop PWR operated by the Pakistan Atomic Energy Commission (PAEC). The radiological source term of CHASNUPP-1 has been evaluated and compared with the advanced modular reactor (SMART) and KORI-1 reactor. For this purpose, modelling and simulation has been carried out in MATLAB. a kinetic model has been developed to carry out the simulation of the release of radionuclides. The core and coolant activity of CHASNUPP-1 is compared with the similar type reactor KORI-1. The developed model uses the ORIGEN 2.2 core inventory as a subroutine. The coolant inventory has been evaluated with 0.25% fuel damage and compared with SMART and KORI-1 reactor.