Decay Heat Uncertainty Analysis for VVER Spent Nuclear Fuel

Kód výsledku v IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F68407700%3A21730%2F25%3A00386159" target="_blank" >RIV/68407700:21730/25:00386159 - isvavai.cz</a>

Výsledek na webu

<a href="https://doi.org/10.1115/1.4068690" target="_blank" >https://doi.org/10.1115/1.4068690</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1115/1.4068690" target="_blank" >10.1115/1.4068690</a>

Jazyk výsledku

angličtina

Název v původním jazyce

Decay Heat Uncertainty Analysis for VVER Spent Nuclear Fuel

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 a standard VVER-440 nuclear fuel. Using the serpent 2 code and referencing the endf/b-viii.0 and tendl nuclear data libraries, the study focused on the impact of cross sections and fission product yield nuclear data uncertainty on the decay heat of a standard VVER-440 nuclear fuel. The time frame chosen for both post-reactor shutdown and long-term for spent fuel cask loading is chosen, providing a comprehensive view of the nuclear fuel cycle.

Název v anglickém jazyce

Decay Heat Uncertainty Analysis for VVER Spent Nuclear Fuel

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 a standard VVER-440 nuclear fuel. Using the serpent 2 code and referencing the endf/b-viii.0 and tendl nuclear data libraries, the study focused on the impact of cross sections and fission product yield nuclear data uncertainty on the decay heat of a standard VVER-440 nuclear fuel. The time frame chosen for both post-reactor shutdown and long-term for spent fuel cask loading is chosen, providing a comprehensive view of the nuclear fuel cycle.

Druh

J_imp - Článek v periodiku v databázi Web of Science

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

2025

Kód důvěrnosti údajů

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

Název periodika

Journal of Nuclear Engineering and Radiation Science

ISSN

2332-8983

e-ISSN

2332-8975

Svazek periodika

11

Číslo periodika v rámci svazku

4

Stát vydavatele periodika

US - Spojené státy americké

Počet stran výsledku

3

Strana od-do

—

Kód UT WoS článku

001572068100001

EID výsledku v databázi Scopus

2-s2.0-105012563929