TOTAL MONTE CARLO UNCERTAINTY ANALYSIS OF VVER-440 SPENT NUCLEAR FUEL IN LWR AND HWR REACTOR ENVIRONMENT

Kód výsledku v IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F68407700%3A21730%2F24%3A00381174" target="_blank" >RIV/68407700:21730/24:00381174 - isvavai.cz</a>

Nalezeny alternativní kódy

RIV/49777513:23220/24:43973889

Výsledek na webu

<a href="https://doi.org/10.1115/ICONE31-136014" target="_blank" >https://doi.org/10.1115/ICONE31-136014</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1115/ICONE31-136014" target="_blank" >10.1115/ICONE31-136014</a>

Jazyk výsledku

angličtina

Název v původním jazyce

TOTAL MONTE CARLO UNCERTAINTY ANALYSIS OF VVER-440 SPENT NUCLEAR FUEL IN LWR AND HWR REACTOR ENVIRONMENT

Popis výsledku v původním jazyce

Total Monte Carlo (TMC) extends the Monte Carlo method, using stochastic techniques and random sampling to solve the Boltzmann transport equation in spent nuclear fuel (SNF) depletion analysis. TMC evaluates the impact of uncertainties in nuclear data, such as cross-sections and fission product yields, on SNF characteristics, focusing on decay heat, which is crucial for SNF handling and management. TMC generates random nuclear data variations within uncertainty ranges for Monte Carlo simulations, resulting in outcome distributions (e.g., decay heat) that reflect real-world behavior uncertainties. The uncertainty analysis examined two reactor environments: a standard VVER-440 reactor using a light water coolant at 1375 MW thermal power and a Teplator district heating reactor VVER-440 fuel in a HWR environment at 50 MW thermal power. This comparison offers insights into VVER-440 SNF behavior under different reactor conditions and coolants. Using the Serpent 2 code and referencing the ENDF/B-VIII.0 and TENDL nuclear data libraries, the study focused on the impact of variable cross-sections on critical nuclides and fission product yield variations. It analyzed decay heat generation immediately post-reactor shutdown and long-term decay heat for spent fuel cask loading, providing a comprehensive view of the nuclear fuel cycle.

Název v anglickém jazyce

TOTAL MONTE CARLO UNCERTAINTY ANALYSIS OF VVER-440 SPENT NUCLEAR FUEL IN LWR AND HWR REACTOR ENVIRONMENT

Popis výsledku anglicky

Total Monte Carlo (TMC) extends the Monte Carlo method, using stochastic techniques and random sampling to solve the Boltzmann transport equation in spent nuclear fuel (SNF) depletion analysis. TMC evaluates the impact of uncertainties in nuclear data, such as cross-sections and fission product yields, on SNF characteristics, focusing on decay heat, which is crucial for SNF handling and management. TMC generates random nuclear data variations within uncertainty ranges for Monte Carlo simulations, resulting in outcome distributions (e.g., decay heat) that reflect real-world behavior uncertainties. The uncertainty analysis examined two reactor environments: a standard VVER-440 reactor using a light water coolant at 1375 MW thermal power and a Teplator district heating reactor VVER-440 fuel in a HWR environment at 50 MW thermal power. This comparison offers insights into VVER-440 SNF behavior under different reactor conditions and coolants. Using the Serpent 2 code and referencing the ENDF/B-VIII.0 and TENDL nuclear data libraries, the study focused on the impact of variable cross-sections on critical nuclides and fission product yield variations. It analyzed decay heat generation immediately post-reactor shutdown and long-term decay heat for spent fuel cask loading, providing a comprehensive view of the nuclear fuel cycle.

Druh

D - Stať ve sborníku

CEP obor

—

OECD FORD obor

20305 - Nuclear related engineering; (nuclear physics to be 1.3);

Projekt

<a href="/cs/project/TN02000012" target="_blank" >TN02000012: Centrum pokročilých jaderných technologií II</a><br>

Návaznosti

P - Projekt vyzkumu a vyvoje financovany z verejnych zdroju (s odkazem do CEP)

Rok uplatnění

2024

Kód důvěrnosti údajů

S - Úplné a pravdivé údaje o projektu nepodléhají ochraně podle zvláštních právních předpisů

Název statě ve sborníku

Proceedings of 31st International Conference on Nuclear Engineering

ISBN

978-0-7918-8822-3

ISSN

—

e-ISSN

—

Počet stran výsledku

8

Strana od-do

—

Název nakladatele

American Society of Mechanical Engineers - ASME

Místo vydání

New York

Místo konání akce

Praha

Datum konání akce

4. 8. 2024

Typ akce podle státní příslušnosti

WRD - Celosvětová akce

Kód UT WoS článku

001349536700050