Short.Term Reactivity Change: Local.Parameter Reactivity Feedback

2021 ◽  
pp. 81-91
Author(s):  
Wei Shen ◽  
Benjamin Rouben
Keyword(s):  
Feedback Effect ◽  
Reactor Operation ◽  
The Core ◽  
Operating Reactor ◽  
Local Parameters ◽  
Reactivity Change

A change in power for an operating reactor generally alters local parameters in the reactor such as the temperatures of the fuel, moderator, and coolant. A change in any of these local parameters causes a change in reactivity that, in turn, affects reactor operation (a feedback effect). Local parameters help to understand the feedback reactivity components related to the core evolution. For the CANDU reactor, the most important local parameters are the following:

New Human-Machine Interface at VR-1 Training Reactor

2010 ◽  
Author(s):  
Martin Kropi´k ◽  
Jan Rataj ◽  
Monika Jurˇicˇkova´
Keyword(s):  
User Interface ◽  
Foreign Students ◽  
Nuclear Power ◽  
English Language ◽  
Reactor Power ◽  
Text Messages ◽  
Reactor Operation ◽  
The Core ◽  
Machine Interface ◽  
And Training

The paper describes a new human-machine (HMI) interface of the VR-1 nuclear training reactor at the Czech Technical University in Prague. The VR-1 reactor is primarily used for training of university students and future nuclear power plant staff. The new HMI was designed to meet functional, ergonomic and aesthetic requirements. It contains a PC with two monitors. The first alphanumerical monitor presents text messages about the reactor operation and status; next, the operator can enter commands to control the reactor operation. The second graphical monitor provides parameters of reactor operation and shows the course of the reactor power and other parameters. Furthermore, it is able to display the core configuration, perform reactivity calculations, etc. The HMI is also equipped with an alarm annunciator. Due to a high number of foreign students and visitors at the reactor, the Czech and English language versions of the user interface are available. The HMI contains also a History server which provides a very detailed storage and future presentation of the reactor operation. The new HMI improves safety and comfort of the reactor utilization, facilitates experiments and training, and provides better support for foreign visitors.

RELAP5-3D Code Application for RBMK-1500 Reactor Core Analysis

2002 ◽  
Author(s):  
Evaldas Bubelis ◽  
Algirdas Kaliatka ◽  
Eugenijus Uspuras
Keyword(s):  
3D Model ◽  
Reactor Core ◽  
Core Analysis ◽  
Reactor Operation ◽  
The Core ◽  
Calculation Results ◽  
Plant Data ◽  
Comparison Of The Results ◽  
Real Plant ◽  
Good Agreement

The paper presents an evaluation of RELAP5-3D code suitability to model specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. A successful best estimate RELAP5-3D model of the Ignalina NPP RBMK-1500 reactor has been developed and validated against real plant data. Certain RELAP5-3D transient calculation results were benchmarked against calculation results obtained using the Russian code STEPAN, specially designed for RBMK reactor analysis. Comparison of the results obtained, using the RELAP5-3D and STEPAN codes, showed quite good mutual coincidence of the calculation results and good agreement with real plant data.

scholarly journals Neutronic Performance of the VVER-1000 Reactor Using Thorium Fuel with ENDF Library

2021 ◽  
Vol 2021 ◽  
pp. 1-9
Author(s):  
Sonia M. Reda ◽  
Ibrahim M. Gomaa ◽  
Ibrahim I. Bashter ◽  
Esmat A. Amin
Keyword(s):  
Uranium Oxide ◽  
Delayed Neutron ◽  
Reactor Operation ◽  
Plutonium Isotopes ◽  
The Core ◽  
Reactivity Coefficients ◽  
Design Values ◽  
Enriched Uranium ◽  
Neutronic Performance ◽  
Lower Production

In this paper, neutronic calculations and the core analysis of the VVER-1000 reactor were performed using MCNP6 code together with both ENDF/B-VII.1 and ENDF/B-VIII libraries. The effect of thorium introduction on the neutronic parameters of the VVER-1000 reactor was discussed. The reference core was initially filled with enriched uranium oxide fuel and then fueled with uranium-thorium fuel. The calculations determine the delayed neutron fraction βeff, the temperature reactivity coefficients, the fuel consumption, and the production of the transuranic elements during reactor operation. βeff and the Doppler coefficient (DC) are found to be in agreement with the design values. It is found that the core loaded with uranium and thorium has lower delayed neutron fraction than the uranium oxide core. The moderator temperature coefficients of the uranium-thorium core are found to be higher than those of the uranium core. Results indicated that thorium has lower production of minor actinides (MAs) and transuranic elements (mainly plutonium isotopes) compared with the relatively large amounts produced from the uranium-based fuel UO2.

Preliminary Evaluation of Core Compaction Phenomenon: Methodological Approach

2014 ◽  
Author(s):  
Rosa Lo Frano ◽  
Giovanni Pugliese ◽  
Giacomo Grasso ◽  
Giuseppe Forasassi
Keyword(s):  
Methodological Approach ◽  
Element Model ◽  
Core Region ◽  
Sensitivity Analyses ◽  
Preliminary Evaluation ◽  
Control Mechanisms ◽  
Reactor Operation ◽  
Core System ◽  
The Core ◽  
Supporting Plate

Nuclear reactors have to be maintained in a critical state so as to keep the chain fission process stationary and under control. Nuclear stability considerations dictate that the geometry of the core be closely controlled at all times: therefore any modification of it must be predictable, compatible with the requirements of the interfacing reactor systems and safely manageable by (intrinsic and engineered) control mechanisms. This study deals with the evaluation of the deformation of core (and restraint system) geometry due to dynamic perturbations. This deformation may determine, at large or small extent, an assembly compaction, that is generally characterized by a radial inward displacement and, eventually, results in an insertion of reactivity. To the aim it is of meaningful importance to set up and develop an overall and reliable methodological approach to be used in designing the core system (all structures and components characterizing the core region) and evaluating its performance under operation and accident condition. In particular a LMR reactor configuration similar to the Advanced Lead-cooled Fast Reactor European Demonstrator - ALFRED (300 MWth) has been considered. The assessment of the dynamic behaviour of a LMR core is particularly needed for seismic design purposes: these solicitations could deform the core system and fuel assembly. A preliminary finite element model, in which all the core sub-assemblies were represented as masses distributed on the supporting plate, was carried out in order to investigate the dynamic response of the structures once confidence was established by sensitivity analyses of size and type of the adopted elements. The preliminary results indicate that the core region is undergoing local deformations (of about 3 cm) that could influence the normal reactor operation. Although any deformation influences the normal reactor operation, it is expected that the reactivity specifically related to this deformation will not pose concerns to the safe manageability of the associated abnormal operation.

scholarly journals Development of a criterion for assessment of fuel washout during operation of WWER power units

2020 ◽  
Vol 6 (4) ◽  
pp. 307-312
Author(s):  
Igor A. Evdokimov ◽  
Andrey G. Khromov ◽  
Petr M. Kalinichev ◽  
Vladimir V. Likhanskii ◽  
Aleksey A. Kovalishin ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Reactor Operation ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Assemblies ◽  
Reliable Detection ◽  
Coolant Activity

Fuel failures may occur during operation of nuclear power plants. One of the possible and most severe consequences of a fuel failure is that fuel may be washed out from the leaking fuel rod into the coolant. Reliable detection of fuel washout is important for handling of leaking fuel assemblies after irradiation is over. Detection of fuel washout is achievable in the framework of coolant activity evaluation during reactor operation. For this purpose, 134I activity is historically used in WWER power units. However, observed 134I activity may increase during operation even if leaking fuel in the core is absent, and fuel deposits are the only source of the fission products release. The paper describes a criterion which enables to reveal the cases when the increase in 134I activity results from the fuel washout from the leaking fuel rods during operation of the WWER-type reactor. Some examples of applications at nuclear power plants are discussed.

scholarly journals Estimation of Fuel Burnup Rate for Core Conversion for the Nigeria Research Reactor-1 (NIRR-1) Fueled With 19.75% Enriched UO2 USING VENTURE PC Code

2020 ◽  
pp. 42-49
Author(s):  
J. A. Rabba ◽  
M. Y. Onimisi ◽  
D. O. Samson
Keyword(s):  
Uranium Dioxide ◽  
Research Reactor ◽  
Fuel Burnup ◽  
Reactor Operation ◽  
The Core ◽  
Computer Codes ◽  
Enriched Uranium ◽  
Conversion Study ◽  
Good Agreement

A standardized burnup analysis using VENTURE-PC computer codes system has been performed for the core conversion study of Nigeria Research Reactor-1. The result obtained from this analysis showed that the mass of Uranium decreases with increase in the number of days of reactor operation while the quantity of Plutonium continues to build up linearly. The buildup of the fissile isotope in the Low Enriched Uranium (LEU) core is very much greater than in the Highly Enriched Uranium (HEU) core. The quantity of Uranium-235 consumed and the amount of Plutonium-239 produce in the core of the reactor were 13.95 g and 0.766745 g respectively for the period of 11 years of reactor operation which is in good agreement with other literatures. This results obtained showed that uranium dioxide (UO2) fuel is a potential material for future Low Enriched Uranium (LEU) core conversion of Nigeria Research Reactor.

Nuclear Analysis During Modification of NPP Core

2010 ◽  
Author(s):  
Andrius Slavickas
Keyword(s):  
Reactor Core ◽  
Reactor Power ◽  
Physical Parameters ◽  
Reactor Operation ◽  
Reactor Control ◽  
The Core ◽  
Burnable Absorber ◽  
Reactivity Coefficients ◽  
Control Rods ◽  
Neutron Activity

Reactor power and neutron activity control is the main key for safe reactor operation. Reactivity coefficients and effects are main measures to estimate reactor control and safety. These characteristics outline reactors behavior during usually exploitation and accident events. Reactivity coefficients and effects quantify the effect, which various parameters (e.g. fuel and graphite temperatures, amount of steam) have for the core neutron activity. Many modifications of RBMK-1500 reactor cores in Ignalina NPP were made during their lifetime. Reactor core modifications like load of higher enriched fuel with burnable absorber and new design control rods affected reactivity coefficients and effects. Neutron-physical parameters calculations of reactor core states with variant fuel loads and new design control rods were performed using QUABOC/CUBBOC-HYCA software. The changes of reactivity coefficients and effects were quantified in this paper.

Regional/Neutron Overpower Protection (ROP/NOP)

2021 ◽  
pp. 59-65
Author(s):  
Wei Shen ◽  
Benjamin Rouben
Keyword(s):  
Neutron Flux ◽  
Reactor Safety ◽  
The Core ◽  
Operating Reactor ◽  
The One ◽  
Reactor Operating ◽  
Adequate Margin

There are independent, separate, and diverse ROP/NOP systems for the two SDS. Each ROP/NOP system consists of a number of flux detectors (see Section 5.4.3) which provide prompt measurements of neutron flux throughout the core. The detectors are mounted inside assemblies that penetrate the core, perpendicular to the fuel channels, vertically or horizontally. The system for SDS1 uses vertical detectors, the one for SDS2 uses horizontal detectors. Detectors are judiciously distributed to monitor the neutron flux throughout the core. In the CANDU 6, a total of 58 ROP detectors are used: 34 for SDS1 and 24 for SDS2. The number and location of detectors in the core are selected in an analysis whose objective is to ensure that as small a number of detectors as possible protect the reactor by tripping a SDS when local high powers threaten reactor safety from any flux shape that could arise in the operating reactor, while at the same time providing adequate margin-to-trip (MTT) to avoid possible restrictions on reactor operating power.

What face familiarity feelings say about the lateralization of specific entities within the core system

2019 ◽  
Vol 42 ◽  
Author(s):  
Guido Gainotti
Keyword(s):  
Temporal Lobe ◽  
Memory System ◽  
Core System ◽  
The Core ◽  
Memory Traces ◽  
Specific Memory ◽  
Target Article ◽  
Temporal Structures ◽  
The Right

Abstract The target article carefully describes the memory system, centered on the temporal lobe that builds specific memory traces. It does not, however, mention the laterality effects that exist within this system. This commentary briefly surveys evidence showing that clear asymmetries exist within the temporal lobe structures subserving the core system and that the right temporal structures mainly underpin face familiarity feelings.

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