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

Result code in 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>

Alternative codes found

RIV/49777513:23220/24:43973889

Result on the web

<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>

Result language

angličtina

Original language name

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

Original language description

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.

Czech name

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Czech description

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Type

D - Article in proceedings

CEP classification

—

OECD FORD branch

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

Project

<a href="/en/project/TN02000012" target="_blank" >TN02000012: Center of Advanced Nuclear Technology II</a><br>

Continuities

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

Publication year

2024

Confidentiality

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

Article name in the collection

Proceedings of 31st International Conference on Nuclear Engineering

ISBN

978-0-7918-8822-3

ISSN

—

e-ISSN

—

Number of pages

8

Pages from-to

—

Publisher name

American Society of Mechanical Engineers - ASME

Place of publication

New York

Event location

Praha

Event date

Aug 4, 2024

Type of event by nationality

WRD - Celosvětová akce

UT code for WoS article

001349536700050