Scram Events Research Articles

Development of a new control rod drive mechanism design for the ISU AGN-201M reactor

The Aerojet General Nucleonics (AGN) model 201-Modified, known as the AGN-201M reactor, plays an essential role in the educational and research activities at Idaho State University (ISU). The ISU AGN-201M's original Control Rod Drive Mechanism (CRDM) has been in operation for more than fifty years with no large-scale redesigns. The CRDM is required to eject the fuel rods within one second during a SCRAM event (also known as a 'reactor trip') and adjust the control rods' insertion speed, and keeps the rod insertion sequence correct. The existing control rod drive mechanisms meet these criteria but experience a few concerns due to aging. Concerns include complex maintenance and costly repairs for old electromechanical components, rod position feedback errors, and the impediment of the plate during a SCRAM due to binding of the lead screws of the existing mechanism. During a binding event, the drive mechanism becomes locked, preventing the control rod's magnetic plate from moving in or out under the reactor's normal and emergency operating conditions. Although, the binding has no effect on the ability of the rod to exit the core during the SCRAM. To counteract the concerns and issues with the current CRDM, a new design has been proposed using newer components and a simplified design.The new design utilizes more advanced electric and mechanical components that are commercially available. The new CRDM system is divided into four main aspects: (1) control rod movement design (motor, lead screw, guide rods), (2) control rod ejection (springs, electromagnet), (3) control rod position and feedback (position transducer, microswitches), and (4) material selection and structural analysis. The new design aims to reduce the overall complexity and probability of failure to improve the reactor's overall reliability. With proper material selection and improved structural design, the new drives are lighter with little to no change in structural integrity. The new control rod drive mechanism eliminates binding scenarios by using a single lead screw and implementing additional guide rods. An advanced linear position sensor and microswitches replace the existing and aging synchro system for accurate rod position feedback resulting in better reactivity control. The new design meets the reactor's operational limits by having an average reactivity insertion of 0.065% Δk/k per second, which corresponds to a total control rod insertion time of 19.23 s, while the control rod's ejection time remains less than one second during a SCRAM event. The new design ensures the reactor's long-term viability for educational and research activities by increasing the reliability and safety of operation for years to come.

Analysis of DAS and ATWS Mitigation System in Nuclear Power Plant

As the backup systems of RPS (reactor protection system), Diverse Actuation System and ATWS( Anticipated transients without scram) mitigation system are diversely designed from RPS system so as to offer protection for the normal operation of nuclear power plant. Both of the systems bear the function of mitigating the anticipated transients without scram events. Being isolated from RPS system, the design of DAS and ATWS system meeting the requirement of diversity and influences of CCFs (common cause failure) are effectively prevented. This article mainly analyzed the structure and function of the two systems and the similarities and differences between them.

Analytical models for a small LFR core dynamics studies

Analytical models for the study of a small Lead-cooled Fast Reactor (LFR) demonstrator (DEMO) core dynamics, in a control-oriented perspective, have been developed aimed at providing a useful, very flexible and straightforward tool allowing relatively quick transient design-basis and stability analyses. A simplified approach has been developed consisting in a lumped-parameter modeling of the coupled neutronics and thermal-hydraulics. The reactor transient responses following both postulated accident initiators such as Unprotected Transient of OverPower (UTOP), Loss of Heat Sink (ULOHS) and Loss of Flow (ULOF), and an emergency SCRAM event have been studied in MATLAB/SIMULINK® environment. A benchmark analysis has been then performed by means of the SAS4A/SASSYS-1 Liquid Metal Reactor Code System with the purpose of providing verification for the analytical outcomes of the nonlinear model and indicating how the latter relate to more realistic one-dimensional calculations. As a general result, responses concerning the main core characteristics (namely, power, reactivity, etc.) have turned out to be mutually consistent in terms of both steady-state absolute figures and transient developments, showing discrepancies of the order of only few per cents. DEMO dynamic behavior has been studied through a linear approach as well, so as to enable the use of linear analysis tools allowing to verify the system stability, thanks to the possibility of expressing a physical model in terms of transfer functions or state–space representation. The impact of linearization and point-kinetics one-precursor-group approximation has been further evaluated by examining the respective transient predictions, which have been compared to the nonlinear reference ones for increasing perturbations. An asymptotic stability analysis, according to Lyapunov's definition, is finally presented.

A Compact Loop-Type Fast Reactor Without Refueling for a Remote Area Power Source

A small reactor has the potential to be utilized as a power source to meet diverse social needs and reduce capital risks. In remote areas, populations tend to be small, and an economic power grid may not be available. In such situations, a small power source with a capacity of less than 50 MW(electric) without refueling is attractive since the costs for fuel transfer to such a site are expensive. In the present study, a metal fuel core with a lifetime of 30 yr and a simple reactor plant design has been proposed. The local burnup reactivity change in every core region is minimized by adjusting the zirconium content and the smear density of the three-core region to achieve a 550°C core outlet temperature. At the end of the cycle, the burnup reactivity is evaluated to be 1.1% of (dk/kk’), achieving a 30-yr core life. The reactor vessel is dramatically simplified by eliminating a fuel-handling system. The number of main cooling loops is reduced to one by installing dual electromagnetic pumps in the primary sodium circuit. The nuclear steam supply system mass, at 309 tonnes, shows that the present loop-type concept can more dramatically reduce material mass than that of the previous pool-type concept of 484 tonnes. The rough estimation of the electricity cost shows that this concept will be competitive for remote sites. Transient analyses show that a self-actuated shutdown system enhances the passive safety features, thus ensuring reactor integrity in anticipated transient without scram events.

3510 1万kWe超小型高速炉(4S炉)の受動安全特性(S49-1 安全,材料,原子炉用機器,S49 原子炉システムおよびその要素技術)

A conceptual design of sodium cooled fast reactor as small-decentralized power supply has been performed. The reactor named 4S (Super Safe, Small and Simple) has a reflector controlled U-Zr metallic fuel core. Characteristics of the 4S reactor are non-refueling (core life time: 30 years), the negative coolant void reactivity during 30 years, integrated and simple reactor structure, and natural circulation decay heat removal systems. Accidents without scram events (AWS) as unprotected sudden loss of flow event and ATWS events have been analyzed for the passive safety features of 4S using a transient simulation code (ARGO-3 code). It is predicts that the passive safety can evade sever damages of the core and primary coolant boundary under the ATWS and AWS events.

Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 °C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 °C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open.

Applicability evaluation to a MOX fueled fast breeder reactor for a self-consistent nuclear energy system

The potential for a MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in a self-consistent nuclear energy system (SCNES). The FBR's annual production amounts of selected long-lived fission products (LLFPs), Se-79, Tc-99, Pd-107, I-129, Cs-135 and Sm-151, can be transmuted by using a two layer radial blanket region without a significant impact on core nuclear and safety characteristics. The other LLFPs are confined in the system. The hazard index level of the LLFPs per one ton of spent fuel from the system after 10 2 years is as small as that of a typical uranium ore. Regarding self-controllability in the system's safety, the proposed FBR core concept has an inherent negative reactivity feedback with a gas expansion module, sodium plenum above the core and burnup reactivity compensation module. So sodium boiling and fuel melting will be avoided in anticipated transient without scram events. Regarding self-terminability, even if the MOX fuel melting should cause a core compaction process, re-criticality of the core can be avoided by a fuel dilution and relocation module.

Mitigation of an Anticipated Transient Without Scram Event in A Simplified Boiling Water Reactor by the Insertion of Fine-Motion Control Rods

Analysis has been performed for an anticipated transient without scram (ATWS) event in a simplified boiling water reactor. Loss of feedwater heating initiated this ATWS, which was mitigated by the insertion of fine-motion control rods. Calculations with RAMONA-4B code show that reactivity insertion, through increase in subcooling to the core, results in a rise in the core power to a new equilibrium condition. Insertion of fine-motion control rods from the bottom of the core is found to shift the peak of the axial power distribution to the upper part of the core until eventual shutdown of the reactor occurs.

Plant Operator Performance Evaluation System

A plant operator performance evaluation system to analyze plant operation records during accident training and to identify and classify operator errors has been developed for the purpose of supporting realization of a training and education system for plant operators. A knowledge engineering technique was applied to evaluation of operator behavior by both event-based and symptom-based procedures, in various situations including event transition due to multiple failures or operational errors The system classifies the identified errors as to their single and double types based on Swain's error classification and the error levels reflecting Rasmussen's cognitive level, and it also evaluates the effect of errors on plant state and then classifies error influence, using “knowledge for phenomena and operations”, as represented by frames. It has additional functions for analysis of error statistics and knowledge acquisition support of “knowledge for operations”. The system was applied to a training analysis for a scram event in a BWR plant, and its error analysis function was confirmed to be effective by operational experts.