Determination of the NPP Krško spent fuel decay heat

Kromar, Marjan; Kurinčič, Bojan

doi:10.1063/1.4994529

Cited by 5 publications

(7 citation statements)

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“…Emissions calculated with the ENDF/B-VII.0 library are almost 40 % lower than emissions predicted with the Origen code. This is somehow in contradiction with the [4] where predicted activities were higher compared to the Origen results. Closer examination has shown that the ENDF/B-VII.0 decay library supplied with the code contains some dubious photon data for some nuclides.…”

Section: Comparison Of the Codes And Librariescontrasting

confidence: 96%

“…Comparison with the Serpent code is performed to verify that the Serpent is taking into account all isotopes important to assess photon and neutron radiation. A similar analysis comparing fuel radioactivity [4] and decay heat [5] has already been performed, showing promising results. After the code validation a sensitivity study is carried out.…”

Section: Introductionmentioning

confidence: 84%

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Assessment of the Photon and Neutron Source Term for the NPP Krško Spent Fuel

Kromar¹,

Kurinčič²

2022

joe

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Accurate knowledge of the fuel nuclide inventory is important after reactor shut down, during the fuel storage and subsequent reprocessing or disposal to provide adequate shielding from the photon and neutron radiation. In this paper possibility to calculate the NPP Krško photon and neutron source term with the Serpent code has been analysed. Some deficiencies in the supplied ENDF/B-VII.0 decay library have been observed. In addition, Serpent reports only spontaneous fission rates without (α, n) and (β, n) contributions. To get neutron emission, spontaneous fission rates had to be multiplied with the average number of neutrons born for each particular nuclide manually. Comparison with the Origen code has shown acceptable agreement of the ENDF/B-VII.1 results. Influence of several factors such as fuel burnup, enrichment, temperature, moderator temperature (density), soluble boron concentration, average power, and burnable absorbers has been analysed. In addition, it was demonstrated that, except for the burnup and enrichment, averaging of all other parameters is acceptable approach. IFBA fuel should be accounted for explicitly due to relative high impact on the photon and neutron emissions.

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Section: Comparison Of the Codes And Librariescontrasting

confidence: 96%

Section: Introductionmentioning

confidence: 84%

Assessment of the Photon and Neutron Source Term for the NPP Krško Spent Fuel

Kromar¹,

Kurinčič²

2022

joe

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“…For decay heat, the calorimetric measurements from CLAB (Swedish Central Intermediate Storage Facility), GE-Morris, and HEDL (Hanford Engineering Development Laboratory) were considered, leading to more than 250 cases. To support the recommendations of this subtask, some studies were already published, see references [3][4][5][6][7][8][9][10][11][12][13][14][15]. A global review of the The aim is to produce NDA systems that can be used to provide experimental data to validate the performance of the codes (including nuclear data) and to verify the declared design and operational history of the assemblies.…”

Section: Sfc Subtask 21: Theoretical Study Of Snf Source Termsmentioning

confidence: 99%

Spent nuclear fuel management, characterisation, and dissolution behaviour: progress and achievement from SFC and DisCo

Sjöland¹,

Christensen²,

Evins³

et al. 2023

EPJ Nuclear Sci. Technol.

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SFC is a work package in Eurad that investigates issues related to the properties of the spent nuclear fuel in the back-end of the nuclear fuel cycle. Decay heat, nuclide inventory, and fuel integrity (mechanical and otherwise), and not least the related uncertainties, are among the primary focal points of SFC. These have very significant importance for the safety and operational aspect of the back-end. One consequence is the operation economy of the back-end, where deeper understanding and quantification allow for significant optimization, meaning that significant parts of the costs can be reduced. In this paper, SFC is described, and examples of results are presented at about half-time of the work package, which will finish in 2024. The DisCo project started in 2017 and finished in November 2021 and was funded under the Horizon 2020 Euratom program. It investigated if the properties of modern fuel types, namely doped fuel, and MOX, cause any significant difference in the dissolution behavior of the fuel matrix compared with standard fuels. Spent nuclear fuel experiments were complemented with studies on model materials as well as the development of models describing the solid state, the dissolution process, and reactive transport in the near field. This research has improved the understanding of processes occurring at the interface between spent nuclear fuel and aqueous solution, such as redox reactions. Overall, the results show that from a long-term fuel matrix dissolution point of view, there is no significant difference between MOX fuel, Cr+Al-doped fuel, and standard fuels.

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“…In (Kromar and Kurinčič, 2017) it was shown that the decay heat is not linearly dependent on burnup and specific power. Therefore, since we have significant burnup and power variations within the fuel assembly in the 3-D approach (Figure 5), the average decay heat is not equal to the decay heat at average burnup and specific power (2-D approach).…”

Section: -D Effectsmentioning

confidence: 99%

“…It is therefore important to have as accurate as possible decay heat calculation tools using minimum number of simplifications that could contribute to the additional uncertainties. In the most widely used approach, the fuel is sliced into one or more independent domains, taking into account some average fuel properties (Kromar and Kurinčič, 2017). Such an approach inevitably leads to discretization uncertainties.…”

Section: Introductionmentioning

confidence: 99%

Determination of the spent fuel decay heat with the VERA core simulator

Kromar

Godfrey

2022

Front. Energy Res.

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VERA core simulator capabilities to predict spent fuel decay heat have been explored. The comparison with the Serpent2 and NEWT/TRITON sequence from the SCALE package, performed on the NPP Krško fuel test cases, has enabled independent verification of the successful implementation of the VERA depletion/decay calculation sequence applied on the extended nuclide set. Since the VERA is a 3-D core simulator, the focus in the rest of the paper was on the investigation of the 3-D effects and non-linearity of the soluble boron concentration averaging process, which are difficult to assess with most of available computational tools. A difference in the decay heat prediction between the 3-D and 2-D model approaches was analyzed. In addition, the impact of the Inconel grids on the decay heat was determined. Averaging the concentration of soluble boron over the fuel irradiation period has been found to be a viable approach, since it results in slight decay heat overprediction, which is considered conservative for most applications. Finally, demonstration on the Watts Bar Unit 1 cycle 1 depletion has shown formidable VERA capabilities to accurately predict spent fuel decay heat.

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Determination of the NPP Krško spent fuel decay heat

Cited by 5 publications

References 1 publication

Assessment of the Photon and Neutron Source Term for the NPP Krško Spent Fuel

Assessment of the Photon and Neutron Source Term for the NPP Krško Spent Fuel

Spent nuclear fuel management, characterisation, and dissolution behaviour: progress and achievement from SFC and DisCo

Determination of the spent fuel decay heat with the VERA core simulator

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