scholarly journals Statistical Burnup Distribution of Moving Pebbles in HTR-PM Reactor

2018 ◽  
Author(s):  
Fanny Vitullo ◽  
Jiri Krepel ◽  
Jarmo Kalilainen ◽  
Horst-Michael Prasser ◽  
Andreas Pautz
Keyword(s):  
Cross Sections ◽  
Fuel Cycle ◽  
Variable Number ◽  
Clear Understanding ◽  
Monte Carlo Code ◽  
The Core ◽  
Average Maximum ◽  
The One ◽  
Under Construction ◽  
Fuel Elements

In the pebble-bed high temperature reactor under construction in China, called HTR-PM, the spherical fuel elements continuously flow downward in the cylindrical core. After the discharge, the burnup of each pebble is checked at the core outlet and, according to the achieved burnup level, the pebble might be disposed or reinserted into the upper section of the core, distributing randomly in the radial direction and defining a variable number of passes necessary to achieve the average maximum burnup of 90 MWd/kgU. Discrete Element Method (DEM) simulations have been carried out to achieve a clear understanding of the movement of 420,000 fuel pebbles in the HTR-PM core. At the same time, neutronic properties have been investigated for a single pebble and for the full core with Serpent 2 Monte Carlo code in order to perform a parametrization of the one-group microscopic cross sections at the core-level. The pebble movement has been coupled with the neutronic behavior of a single pebble in a dedicated burnup script called Moving Pebble Burnup (MPB), developed in Matlab. 3,000 single pebble burnup histories were simulated to obtain sufficient statistics and insight on the burnup process in the HTR-PM. The decrease of the average burnup gained per single pass implies that a miss-handling of recirculated fuel elements is unlikely to lead to exceedance of the maximum allowed burnup of 100 MWd/kgU. Furthermore, the core demonstrates a self-compensation effect of burnup, meaning that it always compensates burnup under- or over-runs in the successive passes. Finally, it is possible to conclude that the fuel cycle of the HTR-PM, as it has been laid out, is well-designed and feasible.

scholarly journals Statistical Burnup Distribution of Moving Pebbles in the High-Temperature Reactor HTR-PM

2020 ◽  
Vol 6 (2) ◽  
Author(s):  
Fanny Vitullo ◽  
Jiri Krepel ◽  
Jarmo Kalilainen ◽  
Horst-Michael Prasser ◽  
Andreas Pautz
Keyword(s):  
High Temperature ◽  
Cross Sections ◽  
Fuel Cycle ◽  
Practical Method ◽  
Clear Understanding ◽  
Monte Carlo Code ◽  
The Core ◽  
High Temperature Reactor ◽  
Under Construction ◽  
Fuel Elements

Abstract In the pebble-bed high-temperature reactor under construction in China, called the HTR-PM, the spherical fuel elements continuously flow downward in the cylindrical core. The burnup of each pebble is checked at the core outlet and, according to the achieved burnup level, the pebble might be disposed or reinserted into the upper section of the core. Upon reinsertion, each pebble is radially distributed in a random manner and, according to its downward path, faces different burnup conditions. Hence, the number of passes necessary to achieve the average discharge burnup of 90 MWd/kgU may vary. Discrete element method (DEM) simulations have been carried out to achieve a clear understanding of the movement of the 420000 fuel pebbles in the HTR-PM core. At the same time, neutronics properties have been investigated for a single pebble and for the full core with the Serpent 2 Monte Carlo code. As a result, one-group microscopic cross sections (XS) have been parametrized at the core level. The pebble movement has been loosely coupled with the depletion of a single pebble in a dedicated burnup script called moving pebble burnup (MPB), developed in matlab. 3000 single pebble burnup histories were simulated to obtain sufficient statistics and an insight into the HTR-PM burnup process. The decrease of the average burnup gained per single pass implies that a miss-handling of recirculated fuel elements is unlikely to lead to an excess of the maximum allowed burnup of 100 MWd/kgU. The core demonstrates a self-compensation effect of burnup, meaning that it always compensates burnup under- or over-runs in the successive passes. In addition, gamma detection of 137Cs has been studied as a practical method for monitoring the burnup of the discharged pebbles, turning out to be an applicable measurement technique. Finally, it is possible to conclude that the fuel cycle of the HTR-PM, as it has been laid out, is well designed and feasible.

Advanced First Core Design for the Westinghouse AP1000

2009 ◽  
Author(s):  
Robert J. Fetterman
Keyword(s):  
Fuel Cycle ◽  
The United States ◽  
Pressurized Water ◽  
Loading Pattern ◽  
The Core ◽  
Low Leakage ◽  
Fuel Assemblies ◽  
Cycle Loading ◽  
Water Reactors ◽  
Under Construction

As the nuclear renaissance is now upon us and new plants are either under construction or being ordered, a considerable amount of attention has also turned to the design of the first fuel cycle. Requirements for core designs originate in the Utilities Requirements Document (URD) for the United States and the European Utilities Requirements (EUR) for Europe. First core designs created during the development of these documents were based on core design technology dating back to the 1970’s, where the first cycle core loading pattern placed the highest enrichment fuel on the core periphery and two other lower enrichments in the core interior. While this sort of core design provided acceptable performance, it underutilized the higher enriched fuel assemblies and tended to make transition to the first reload cycle challenging, especially considering that reload core designs are now almost entirely of the Low Leakage Loading Pattern (LLLP) design. The demands placed on today’s existing fleet of pressurized water reactors for improved fuel performance and economy are also desired for the upcoming Generation III+ fleet of plants. As a result of these demands, Westinghouse has developed an Advanced First Core (AFCPP) design for the initial cycle loading pattern. This loading pattern design simulates the reactivity distribution of an 18 month low leakage reload cycle design by placing the higher enriched assemblies in the core interior which results in improved uranium utilization for those fuel assemblies carried through the first and second reload cycles. Another feature of the advanced first core design is radial zoning of the high enriched assemblies, which allows these assemblies to be located in the core interior while still maintaining margin to peaking factor limits throughout the cycle. Finally, the advanced first core loading pattern also employs a variety of burnable absorber designs and lengths to yield radial and axial power distributions very similar to those found in typical low leakage reload cycle designs. This paper will describe each of these key features and demonstrate the operating margins of the AFC design and the ability of the AFC design to allow easy transition into 18 month low leakage reload cycles. The fuel economics of the AFC design will also be compared to those of a more traditional first core loading pattern.

Characterization of Radioactive Waste From Side Structural Components of a CANDU Reactor for Decommissioning Applications in Korea

2010 ◽  
Author(s):  
Rizwan Ahmed ◽  
Gyunyoung Heo ◽  
Dong-Keun Cho ◽  
Jongwon Choi
Keyword(s):  
Radioactive Waste ◽  
Cross Sections ◽  
Reactor Core ◽  
International Standards ◽  
Monte Carlo Code ◽  
Capture Cross ◽  
Structural Components ◽  
The Core ◽  
Waste Classification ◽  
Capture Cross Sections

Reactor core components and structural materials of nuclear power plants to be decommissioned have been irradiated by neutrons of various intensities and spectrum. This long term irradiation results in the production of large number of radioactive isotopes that serve as a source of radioactivity for thousands of years for future. Decommissioning of a nuclear reactor is a costly program comprising of dismantling, demolishing of structures and waste classification for disposal applications. The estimate of radio-nuclides and radiation levels forms the essential part of the whole decommissioning program. It can help establishing guidelines for the waste classification, dismantling and demolishing activities. ORIGEN2 code has long been in use for computing radionuclide concentrations in reactor cores and near core materials for various burn-up-decay cycles, using one-group collapsed cross sections. Since ORIGEN2 assumes a constant flux and nuclide capture cross-sections in all regions of the core, uncertainty in its results could increase as region of interest goes away from the core. This uncertainty can be removed by using a Monte Carlo Code, like MCNP, for the correct calculations of flux and capture cross-sections inside the reactor core and in far core regions. MCNP has greater capability to model the reactor problems in much realistic way that is to incorporate geometrical, compositional and spectrum information. In this paper the classification of radioactive waste from the side structural components of a CANDU reactor is presented. MCNP model of full core was established because of asymmetric structure of the reactor. Side structural components of total length 240 cm and radius 16.122 cm were modeled as twelve (12) homogenized cells of 20 cm length each along the axial direction. The neutron flux and one-group collapsed cross-sections were calculated by MCNP simulation for each cell, and then those results were applied to ORIGEN2 simulation to estimate nuclide inventory in the wastes. After retrieving the radiation level of side structural components of in- and ex-core, the radioactive wastes were classified according to the international standards of waste classification. The wastes from first and second cell of the side structural components were found to exhibit characteristics of class C and Class B wastes respectively. However, the rest of the waste was found to have activity levels as that of Class A radio-active waste. The waste is therefore suitable for land disposal in accordance with the international standards of waste classification and disposal.

scholarly journals Study of the jet stream over India and to its north in winter: Part I

MAUSAM ◽  
2021 ◽  
Vol 22 (1) ◽  
pp. 1-14
Author(s):  
SQN. LDR. M. S. SINGH
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Polar Front ◽  
Maximum Speed ◽  
Jet Stream ◽  
Jet Streams ◽  
Model Cross Section ◽  
Average Maximum ◽  
The One ◽  
Polar Front Jet

Characteristics of the jet streams over India and to its north in winter were studied with the daily vertical cross sections (1200 GMT) along 75°E from 8oN to 60°N for the period I to 15 February 1967, It was observed that there are three separate jet cores present in this latitl1de belt on most of the days, located on an average at 43°N, 31°N and 23°N. of these three, the most stable and persistent one is the second which is located between Delhi and Srinagar, at 200-mb level with an average maximum speed of 140-150 kt. The one to its south is weaker and quite variable in location as well as altitude. The jet at 31°N, therefore, has been called the primary sl1b-tropical jet over India and its characteristics studied. Based on this study, a. model cross-section has been. prepared for this STJ, The descriptions of the STJ at 23°N and also of PFJ (Polar Front Jet) at 43°N are included.

Neutronic Analysis for Tehran Research Reactor Upgrading

2002 ◽  
Author(s):  
Ebrahim Afshar ◽  
Alireza Shahidi
Keyword(s):  
Research Reactor ◽  
Three Dimensional ◽  
Operating Conditions ◽  
Monte Carlo Code ◽  
Radioisotope Production ◽  
The Core ◽  
Transport Code ◽  
Tehran Research Reactor ◽  
Fuel Elements ◽  
Good Agreement

This paper presents preliminary results of a study undertaken to investigate the possibility of raising the power of Tehran Research Reactor (TRR) from 5 to 10 MW (th), keeping the same core configuration and with minimum changes in the primary cooling circuit. The main aim of TRR upgrade is to increase the volume of radioisotope production. The neutronic analysis was carried out for a fresh core with 22 Standard Fuel Elements (SFE) under normal operating conditions. Two different calculational lines were used to simulate the neutronic behavior in the core and perform the necessary neutronic calculations. First, combination of cell calculation transport code WIMS-D/4 [1] and three dimensional core calculation diffusion code CITATION [2] were used to and next a Monte Carlo code MCNP-4B [3] together with point depletion code ORIGEN-2 [4] were used. The results obtained show good agreement between these two different schemes.

Fuel Assembly Bowing and Core Restraint Design in Fast Reactors

2014 ◽  
Author(s):  
J. J. Grudzinski ◽  
C. Grandy
Keyword(s):  
Fuel Assembly ◽  
Cross Sections ◽  
Fast Reactor ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Transient Behavior ◽  
Restraint System ◽  
The Core ◽  
Fuel Elements ◽  
Over Time

The reactivity of a fast spectrum nuclear reactor core is sensitive to changes in the fuel position. The core is formed by a hexagonal array of fuel assemblies which contain the fuel elements. The main structural components of the assemblies are thinwall hexagonal ducts. The fuel elements represent negligible stiffness in the fuel assembly compared to the ducts such that the ducts determine the location of the fuel. Thermal gradients across the fuel assembly cross sections create differential thermal expansion which causes the assemblies to bow. This bowing is proportional to the power to flow ratio such that it can become an important part of the reactivity change during reactor transients such as during reactor start-up, transient overpower (TOP), and unprotected loss of flow without scram (ULOF). In addition to these short term transients, thermal and fast neutron flux gradients within the core cause the assembly ducts to swell and bow over time due to irradiation creep and swelling. These latter effects produce permanent bowing of the ducts which change the reactivity over time and more importantly affect the mechanical forces required to refuel the core as the bowing is greater that the duct-to-duct clearance. Understanding these bowing responses is important to the understanding of both the transient behavior of a fast reactor as well as the refueling loads. Through proper design of the core restraint system, the bowing response can be engineered to provide negative feedback during the above mentioned transients such that it becomes part of the inherent safety of a fast reactor. Similarly, the opposing effects of creep and swelling can be manipulated to reduce the permanent core bowing deformations. We provide a discussion of the key features of analyzing and designing a core restraint system and provide a brief survey of the past work.

The Salvage of Ideas: Schumpeter and the Core Problems Still Confronting Economics

2003 ◽  
pp. 15-26
Author(s):  
P. Wynarczyk
Keyword(s):  
The Other ◽  
The Core ◽  
Other Hand ◽  
The One

Two aspects of Schumpeter' legacy are analyzed in the article. On the one hand, he can be viewed as the custodian of the neoclassical harvest supplementing to its stock of inherited knowledge. On the other hand, the innovative character of his works is emphasized that allows to consider him a proponent of hetherodoxy. It is stressed that Schumpeter's revolutionary challenge can lead to radical changes in modern economics.

AFRICAN WOMEN AND FORBIDDEN GROUNDS: FEMALE SEXUALITY AND SELF-DETERMINATION IN AFRICAN LITERATURE

Imbizo ◽  
2017 ◽  
Vol 7 (1) ◽  
pp. 40-54
Author(s):  
Oyeh O. Otu
Keyword(s):  
Female Sexuality ◽  
Sense Of Self ◽  
African Literature ◽  
The Other ◽  
Self Determination ◽  
African Women ◽  
The Core ◽  
The Many ◽  
The One ◽  
Sexual Repression

This article examines how female conditioning and sexual repression affect the woman’s sense of self, womanhood, identity and her place in society. It argues that the woman’s body is at the core of the many sites of gender struggles/ politics. Accordingly, the woman’s body must be decolonised for her to attain true emancipation. On the one hand, this study identifies the grave consequences of sexual repression, how it robs women of their freedom to choose whom to love or marry, the freedom to seek legal redress against sexual abuse and terror, and how it hinders their quest for self-determination. On the other hand, it underscores the need to give women sexual freedom that must be respected and enforced by law for the overall good of society.

Evaluation of Road Utilisation as Space For Holding Social Ceremonies in Indigenous Residential Area of Ogbomoso, Nigeria

2018 ◽  
Vol 1 (March 2018) ◽  
Author(s):  
S.A Okanlawon ◽  
O.O Odunjo ◽  
S.A Olaniyan
Keyword(s):  
Transport Infrastructure ◽  
Noisy Environment ◽  
Policy Makers ◽  
The Core ◽  
Local Roads ◽  
Strict Enforcement ◽  
The Mean ◽  
The One ◽  
The City ◽  
The Town

This study examined Residents’ evaluation of turning transport infrastructure (road) to spaces for holding social ceremonies in the indigenous residential zone of Ogbomoso, Oyo State, Nigeria. Upon stratifying the city into the three identifiable zones, the core, otherwise known as the indigenous residential zone was isolated for study. Of the twenty (20) political wards in the two local government areas of the town, fifteen (15) wards that were located in the indigenous zone constituted the study area. Respondents were selected along one out of every three (33.3%) of the Trunk — C (local) roads being the one mostly used for the purpose in the study area. The respondents were the residents, commercial motorists, commercial motorcyclists, and celebrants. Six hundred and forty-two (642) copies of questionnaire were administered and harvested on the spot. The Mean Analysis generated from the respondents’ rating of twelve perceived hazards listed in the questionnaire were then used to determine respondents’ most highly rated perceived consequences of the practice. These were noisy environment, Blockage of drainage by waste, and Endangering the life of the sick on the way to hospital; the most highly rated reasons why the practice came into being; and level of acceptability of the practice which was found to be very unacceptable in the study area. Policy makers should therefore focus their attention on strict enforcement of the law prohibiting the practice in order to ensure more cordial relationship among the citizenry, seeing citizens’ unacceptability of the practice in the study area.

scholarly journals Method for assessment of matching of total ankle joint endoprostheses based on CT data

2020 ◽  
Vol 6 (3) ◽  
pp. 396-397
Author(s):  
Heiner Martin ◽  
Josephine Wittmüß ◽  
Thomas Mittlmeier ◽  
Niels Grabow
Keyword(s):  
Cross Sections ◽  
Spline Functions ◽  
Moments Of Inertia ◽  
Cad System ◽  
Principal Moments Of Inertia ◽  
Principal Axes ◽  
Ct Data ◽  
The One ◽  
Relative Differences ◽  
Systemic Design

AbstractThe investigation of matching of endoprosthesis tibial components to the bone cross section is of interest for the manufacturer as well as for the surgeon. On the one hand, a systemic design of the prosthesis and the assortment is possible, on the other hand, a better matching implantation is enabled on the basis of experience of this study. CT sections were segmented manually using a CAD system and fitted by spline functions, then superseded with cross sections of the tibial component of a modified Hintermann H3 prosthesis. The principal moments of inertia, the direction of the principal axes and the area of the section were evaluated. Based on the relative differences of the principal moments of inertia, recommendations for application of the different prosthesis size and its selection with the surgery can be made.

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