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Whole core burnup calculations using `MCNP`

Abstract

Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It  More>>
Authors:
Haran, O; Shaham, Y [1] 
  1. Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev
Publication Date:
Dec 01, 1996
Product Type:
Conference
Report Number:
INIS-mf-15507; CONF-961252-
Reference Number:
SCA: 210100; PA: AIX-28:023734; EDB-97:072306; SN: 97001788837
Resource Relation:
Conference: 19. conference of the Israel Nuclear Societies, Herzliya (Israel), 9-10 Dec 1996; Other Information: PBD: Dec 1996; Related Information: Is Part Of Program and book of abstracts; PB: 149 p.
Subject:
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; BURNUP; M CODES; ALGORITHMS; BWR TYPE REACTORS; DATA-FLOW PROCESSING; NUCLEAR FUELS
OSTI ID:
475952
Research Organizations:
Israel Nuclear Society, Yavne (Israel)
Country of Origin:
Israel
Language:
English
Other Identifying Numbers:
Other: ON: DE97616259; TRN: IL9606205023734
Availability:
INIS; OSTI as DE97616259
Submitting Site:
INIS
Size:
pp. 1-4
Announcement Date:

Citation Formats

Haran, O, and Shaham, Y. Whole core burnup calculations using `MCNP`. Israel: N. p., 1996. Web.
Haran, O, & Shaham, Y. Whole core burnup calculations using `MCNP`. Israel.
Haran, O, and Shaham, Y. 1996. "Whole core burnup calculations using `MCNP`." Israel.
@misc{etde_475952,
title = {Whole core burnup calculations using `MCNP`}
author = {Haran, O, and Shaham, Y}
abstractNote = {Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors).}
place = {Israel}
year = {1996}
month = {Dec}
}