scholarly journals COMPARISON OF WASTE DUE TO IRRADIATED STEELS IN THE ESFR AND DEMO

2021 ◽  
Vol 247 ◽  
pp. 18002 ◽  
Author(s):  
Jack Reid ◽  
Greg Bailey ◽  
Edmund Cracknell ◽  
Mark Gilbert ◽  
Lee Packer
Keyword(s):  
Energy Source ◽  
Waste Disposal ◽  
Fusion Reactor ◽  
Neutron Transport ◽  
Intermediate Level ◽  
Lower Percentage ◽  
The Core ◽  
Simulation Process ◽  
Fission Reactors ◽  
Reduced Activation Steels

For either nuclear fusion or generation IV fission reactors to be viable as a commercial energy source the decommissioning and waste disposal solutions must be considered during the design. A multi-step simulation process combining Monte Carlo Neutron Transport simulations with inventory simulations have been performed to estimate the activation of steels in key reactor components of the European Sodium-cooled fast Reactor (ESFR). Waste classifications based on UK waste disposal regulations have been applied to the key components to estimate the expected masses of low level and intermediate level waste. The use of reduced activation steels, EUROFER and F82H, in reactor components external to the core results in a factor of 10 reduction in the percentage of waste classified as Intermediate Level Waste (ILW). Waste estimates are compared to existing waste estimates for the European Demonstration fusion reactor (DEMO). The ESFR has a lower percentage of ILW per total reactor mass due to irradiated steels compared to DEMO. However, there is no Higher Activity Waste (HAW) associated with DEMO, compared with arisings from the ESFR spent fission fuel.

scholarly journals Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

2020 ◽  
Vol 225 ◽  
pp. 03007
Author(s):  
Tanja Goričanec ◽  
Domen Kotnik ◽  
Žiga Štancar ◽  
Luka Snoj ◽  
Marjan Kromar
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Reactor Core ◽  
Response Prediction ◽  
Detector Response ◽  
Design Calculation ◽  
Monte Carlo Code ◽  
Control Rod ◽  
The Core ◽  
Fuel Assemblies

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

Development of nuclear criticality safety controls on intermediate-level waste packages

2012 ◽  
Vol 76 (8) ◽  
pp. 2949-2956 ◽  
Author(s):  
T. W. Hicks ◽  
P. Wood ◽  
D. Putley ◽  
T. D. Baldwin
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Neutron Transport ◽  
Intermediate Level ◽  
Fissile Material ◽  
Radioactive Waste Management ◽  
Waste Package ◽  
Operational Phase ◽  
Nuclear Criticality Safety ◽  
Enriched Uranium

AbstractIntermediate-level wastes (ILW) include substantial quantities of fissile material and controls are required to ensure that its storage, transport and disposal does not present a nuclear criticality hazard. This paper describes the Radioactive Waste Management Directorate's research to develop package fissile material limits (in the form of screening levels) for four different categories of ILW, defined according to uranium or plutonium composition: (1) irradiated natural and slightly enriched uranium (uranium containing up to 1.9 wt.% 235U); (2) low-enriched uranium (uranium containing up to 4 wt.% 235U); (3) high-enriched uranium (uranium containing up to 100 wt.% 235U); and (4) separated plutonium (plutonium containing up to 100 wt.% 239Pu).The derivation of package screening levels was supported by neutron transport calculations that addressed conditions during waste package transport to a geological disposal facility (GDF), during the GDF operational phase and after GDF closure. The analysis included consideration of combinations of events and processes that could result in fissile material accumulation and concentration after GDF closure, when waste packages have deteriorated sufficiently for fissile material to be mobilized. The results of the calculations have provided input to Radioactive Waste Management Directorate's decision making on setting waste package screening levels.

Large-Scale Model Testing of a Concrete Silo EBS for Low- and Intermediate-Level Waste Disposal

1997 ◽  
Vol 506 ◽  
Author(s):  
K. Konishi ◽  
K. Ukaji ◽  
A. Fujiwara ◽  
T. Endo ◽  
Y. Tsuji
Keyword(s):  
Thermal Stress ◽  
Radioactive Waste ◽  
Waste Disposal ◽  
Large Scale ◽  
Heat Of Hydration ◽  
Intermediate Level ◽  
Scale Model ◽  
Large Scale Model ◽  
One Year ◽  
Intermediate Level Radioactive Waste

ABSTRACTThis study focuses on demonstrating the feasibility of constructing a large reinforced concrete silo for the disposal of low- and intermediate-level radioactive waste. In a large-scale model test, the results of temperature (from heat of hydration) and thermal stress analyses were confirmed, as was the practicality of countermeasures to prevent thermal cracking. After one year, the average permeability of the model silo was estimated to be below 1×10−11m/s.

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