Safety Rods Research Articles

Analysis of unprotected transients with control and safety rod drive mechanism expansion feedback in a medium sized oxide fuelled fast breeder reactor

Possibilities of enhancing core safety under unprotected loss of flow (ULOF) and unprotected transient over power (UTOP) accidents with control and safety rod drive mechanism (CSRDM) expansion feedbacks are explored in a medium sized oxide fuelled fast breeder reactor. This feedback is expected to take the reactor to a safe shutdown under ULOF and to an another steady state under UTOP where there is no significant fuel melting. Under ULOF, with CSRDM feedback net reactivity was maintained negative throughout the transient (up to 2000s) and the power dropped to a level of heat removal capacity of decay heat removal system based on natural circulation. Similarly, under UTOP with the above feedback reactor power goes to a lower peak value. The fuel temperature is just touching the melting temperature and the melt fraction does not cross 5%. With CSRDM expansion feedbacks both ULOF and UTOP transients prolong beyond 2000s. It ensures, availability of time for initiating any safety actions against the transients, and thus it helps to preclude core disruptive accidents (CDA) in a medium sized oxide fuelled reactors.Classification: L. safety and risk analysis

Распухание направляющих гильз стержней СУЗ быстрых реакторов в неоднородных температурных и радиационных полях

Распухание направляющих гильз стержней СУЗ быстрых реакторов в неоднородных температурных и радиационных полях

Level monitoring system with pulsating sensor--application to online level monitoring of dashpots in a fast breeder reactor.

An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

Experimental studies on CSRDM specimen to validate the use of Ritz vectors for the response evaluation of structures subjected to multi-support excitation

The control and safety rod drive mechanism is a classic example of a multiply supported and differentially excited structure as it receives varying inputs at each of its supports as they are connected at different parts of a building. The responses of such systems are conventionally computed using pseudo-static method in which the dynamic component is evaluated by time history analysis, response spectrum method or modal analysis. In the present work the dynamic components are evaluated by using Ritz vectors and are compared with the conventional methods namely time history analysis and modal analysis. The usage of Ritz vectors as an alternative for conventional modal vectors has been validated through an experiment conducted on a control and safety rod drive mechanism (CSRDM) subjected to differential input motion. The validation of the code developed for this purpose is carried out for a bridge structure.

Light and heavy water dynamic structure factor for neutron transport codes

In this study, we report on recent neutron inelastic scattering experiments performed at the Institut Laue-Langevin (ILL) for H2O and D2O. The measured dynamic structure factors S(q, ω) have been reduced, normalised and transformed into the S(α, β) formalism, where α and β stand for the unit-less momentum and energy transfers, respectively. The measurements were complemented with molecular dynamics simulations. After processing with NJOY, new water neutron scattering cross-sections have been generated for use with e.g. the Monte Carlo N-Particle (MCNP) software in view to improve the accuracy of the nuclear facility models. As an example, we present improved accuracy calculations for the safety rod insertion impact on the criticality factor keff for the ILL high flux research reactor.

Mathematical modelling of performance of safety rod and its drive mechanism in sodium cooled fast reactor during scram action

Performance of safety rod and its drive mechanism which are parts of shutdown systems in sodium cooled fast reactor (SFR) plays a major role in ensuring safe operation of the plant during all the design basis events. The safety rods are to be inserted into the core within a stipulated time during off-normal conditions of the reactor. Mathematical modelling of dynamic behaviour of a safety rod and its drive mechanism in a typical 500MWe SFR during scram action is considered in the present study. A full-scale prototype system has undergone qualification tests in air, water and in sodium simulating the operating conditions in the reactor. In this paper, the salient features of the safety rod and its mechanism, details related to mathematical modelling and sensitivity of the parameters having influence on drop time are presented. The outcomes of the numerical analysis are compared with the experimental results. In this process, the mathematical model and the computer software developed are validated.

Assessment of sensitivity of neutron-physical parameters of fast neutron reactor to purification of reprocessed fuel from minor actinides

The work is devoted to computational investigation of the dependence of basic physical parameters of fast neutron reactors on the degree of purification of plutonium from minor actinides obtained as a result of pyroelectrochemical reprocessing of spent nuclear fuel and used for manufacturing MOX fuel to be reloaded into the reactors mentioned. The investigations have shown that, in order to preserve such important parameters of a BN-800 type reactor as the criticality, the sodium void reactivity effect, the Doppler effect, and the efficiency of safety rods, it is possible to use the reprocessed fuel without separation of minor actinides for refueling (recharging) the core.

On the interpretation of the inverted kinetics equation and space-time calculations of the effectiveness of the VVER-1000 reactor scram system

In the present paper, an attempt is made to analyze the accuracy of calculating the effectiveness of the VVER-1000 reactor scram system by means of the inverted solution of the kinetics equation (ISKE). In the numerical studies in the intellectual ShIPR software system, the actuation of the reactor scram system with the possible jamming of one of the two most effective rods is simulated. First, the connection of functionals calculated in the space-time computation in different approximations with the kinetics equation is considered on the theoretical level. The formulas are presented in a manner facilitating their coding. Then, the results of processing of several such functions by the ISKE are presented. For estimating the effectiveness of the VVER-1000 reactor scram system, it is proposed to use the measured currents of ionization chambers (IC) jointly with calculated readings of IC imitators. In addition, the integral of the delayed neutron (DN) generation rate multiplied by the adjoint DN source over the volume of the reactor, calculated for the instant of time when insertion of safety rods ends, is used. This integral is necessary for taking into account the spatial reactivity effects. Reasonable agreement was attained for the considered example between the effectiveness of the scram system evaluated by this method and the values obtained by steady-state calculations as the difference of the reciprocal effective multiplication factors with withdrawn and inserted control rods. This agreement was attained with the use of eight-group DN parameters.

3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs) is made along with a design change in CSRs to increase the reactivity worth of absorber rods for ensuring safe reactor operation. This new enhanced burnup core can be used in the future with advanced structural materials for clad and hexcan, which can withstand higher neutron fluence and gamma radiation.

Design of instrumentation for control and safety rod drive mechanisms of prototype fast breeder reactor

Prototype Fast Breeder Reactor (PFBR) is a 500 MWe, sodium cooled, plutonium oxide-uranium oxide fuelled, pool type fast breeder nuclear reactor under construction in India. Control and Safety Rod Drive Mechanism (CSRDM) is part of one of the shutdown systems and is used to drive neutron absorber rods up and down. Instrumentation and control for CSRDM realises significant functions like power control, power set back and controlled shutdown. A study was carried out into incidents during the operation of 40 MWt experimental reactor called Fast Breeder Test Reactor and testing of a prototype CSRDM. Significant incidents in FBTR like uncontrolled withdrawal of absorber rod, movement in unintended direction, partial absorber rod drops are found to have occurred. Corrective measures are taken in the I&C of PFBR. PFBR I&C for CSRDM is presented. I&C designers for nuclear systems may gainfully make use of the experiences described in the paper.

Experimental Results From the VENUS-F Critical Reference State for the GUINEVERE Accelerator Driven System Project

The GUINEVERE (Generation of Uninterrupted Intense NEutron pulses at the lead VEnus REactor) project was launched in 2006 within the framework of FP6 EUROTRANS in order to validate online reactivity monitoring and subcriticality level determination in accelerator driven systems (ADS). Therefore, the VENUS reactor at SCK-CEN in Mol, Belgium, was modified towards a fast core (VENUS-F) and coupled to the GENEPI-3C accelerator built by CNRS. The accelerator can operate in both continuous and pulsed mode. The VENUS-F core is loaded with enriched Uranium and reflected with solid lead. A well-chosen critical reference state is indispensable for the validation of the online subcriticality monitoring methodology. Moreover, a benchmarking tool is required for nuclear data research and code validation. In this paper, the design and the importance of the critical reference state for the GUINEVERE project are motivated. The results of the first experimental phase on the critical core are presented. The control rods worth is determined by the positive period method and the application of the Modified Source Multiplication (MSM) method allows the determination of the worth of the safety rods. The results are implemented in the VENUS-F core certificate for full exploitation of the critical core.

Flow induced vibration studies on PFBR control plug components

The construction of Prototype Fast Breeder Reactor (PFBR), a 500MWe liquid sodium cooled reactor, is in progress at Kalpakkam in India. Control plug (CP) is located right above the core subassemblies in the hot pool. Control plug is an important component as many of the critical reactor parameters are sensed and controlled by the components housed in the control plug assembly. In PFBR primary circuit, components are basically thin walled, slender shells with diameter to thickness ratio ranging from 100 to 650. These components are prone to flow induced vibrations. The existence of free liquid (sodium) surfaces, which is the source of sloshing phenomenon and the operation of primary sodium pump in the primary pool are other potential sources of vibration of reactor components.Control plug is a hollow cylindrical shell structure and provides passages and support for 12 absorber rod drive mechanisms (ARDM) which consists of 9 control and safety rods and 3 diverse safety rods, 210 thermo wells to measure the sodium temperature at the exit of various fuel subassemblies, three failed fuel localization modules (FFLM) and acoustic detectors. It consists of a core cover plate (CCP), which forms the bottom end, two intermediate supports plate, i.e. lower stay plate (LSP) and upper stay plate (USP) and an outer shell. The CCP is located at a distance of 1.3m from the core top. With such a gap, there will be long free hanging length of the thermocouple sleeves, Delayed neutron detector (DND) sampling tubes and ARDM shroud tubes and hence they are susceptible to flow induced vibrations. Since control plug is partially immersed in hot sodium, the reactor transients are felt by the components, hence it is very much essential to understand the vibration response of the control plug components. The main vibration excitation mechanisms are vortex shedding, turbulence buffeting, fluid-elastic instability, etc.In order to assess the susceptibility of CP internals against flow induced vibrations (FIVs), to measure structural response and to validate the analytical codes developed for FIV analysis of CP along with the assumptions of supports for various CP internals therein, a flow induced vibration testing program was formulated in 1:4 scale model of PFBR CP. As the first phase of this program, experimental modal analysis of CP internals was carried out in air to estimate the modal parameters. Subsequently, flow induced vibration studies were conducted in the 1:4 scale model of CP in water. In this model, size of the tubes, shell and plates are reduced to ¼ size of the PFBR CP. The stiffness of CP parts is reduced by 4 times and mass decreased by 64 times which results in scaling up of modal frequencies by 4 times. The CP internals in the model were instrumented with accelerometers and strain gages. The studies were conducted in water with flows derived on the basis of velocity similitude. The output signals from the sensors were acquired and analyzed to obtain frequency spectra, overall vibration amplitude and strain values at various locations inside CP. The study carried out confirms the absence of resonance due to flow induced vibration mechanisms for the entire operating range. This paper elucidates the modeling details, similitude criteria, instrumentation employed and experimental results obtained with discussion on results.

Development of sodium leak detectors for PFBR

The 500MWe Prototype Fast Breeder Reactor (PFBR) is under advanced stage of construction at Kalpakkam near Chennai in India. The wide and high operating temperature, highly chemically active nature of sodium and its reaction with air make the sodium instrumentation complex over the conventional instrumentation. Over the years, traditional instruments such as wire type leak detectors, spark plug type leak detectors were developed and used in different sodium systems. The redundant and diverse leak detection method calls for development of special instrumentation for sodium systems which include sodium ionization (leak) detector for detecting minute sodium leak in addition to those systems based on mutual inductance principle. For detection of sodium leak from reactor Main Vessel (MV), diverse methods are used such as miniature, remotely locatable, Mutual Inductance type Leak Detector(MILD) and specially modified spark plug type leak detector. The design of MILD is suitably modified for detecting leak in double wall pipes and Diverse Safety Rod drive Mechanism (DSRDM). Steam/water leak in steam generator produces hydrogen and leads to high pressure and temperature in the system. Rupture disc is used as a safety device which punctures itself due to sudden pressure rise. To detect the discharge of sodium and its reaction products at the downstream of the rupture disc due to bursting of the rupture disc, cross wire type leak detector has been designed, developed and tested. The selection of the leak detection system depends on the location where leak has to be detected. This paper describes the recent developments carried out in leak detector systems to meet the challenges involved in developing robust sensors for PFBR including miniaturization of leak detectors and the ability to locate or withdraw them remotely through a long guide tube with complex profile.

Sensor development for position detection of diverse safety rod of fast reactor

The 500MWe sodium cooled prototype fast breeder reactor (PFBR) presently under advanced stage of construction has two diverse and independent shut-down systems. Each shutdown system comprises of independent sensing systems, control logic and drive mechanisms. The drive mechanism belonging to the first shutdown system is used for controlling reactor power and also for shutting down the reactor. It is known as control and safety drive rod mechanism (CSRDM). The mechanism belonging to the second shutdown system is used only for shutting down the reactor and is called diverse safety rod drive mechanism (DSRDM). CSRDM has potentiometer which is used to determine the position of the control and safety rod (CSR). The free fall time of control and safety rod can also be measured with the help of the potentiometer. Constraint of putting similar type of instrument in DSRDM has prohibited position detection of DSR as well as measurement of free fall time of diverse safety rod (DSR). However center has embarked on development of three different techniques to detect the position and free fall time of DSR. These techniques are Kalman filtering technique using neutronic signal to detect position of DSR, measurement of acoustic signal which is produced when DSR hits the dashpot and the eddy current based technique. Description of development of eddy current based sensor to detect position and free fall time of DSR is the main aim of this paper. The sensor consists of five coils which are inductively coupled. The change in position of DSR causes change in inductance of the coil and hence voltage induced in the coil. The measurement of coil voltage reveals the position of the DSR. Sodium testing of the sensor has been carried out on a 1:1 model of the sensor. The performance of the sensor at various temperatures and frequencies has been studied and found satisfactory. The sodium testing has shown encouraging results and have proven the feasibility of using such a sensor for the reactor application.The piston of the DSR is made of modified 9Cr–1Mo magnetic steel. A sensor winding is placed around the dashpot, the inductance of the winding changes when the DSR is deposited in the dashpot. Transformer coupling between two windings is then used to measure the inductance change of the sensor winding to facilitate non-contact measurement. The winding configuration has been chosen such that the signal can be picked-up even when there is misalignment between the coils due to thermal and radiation induced dimensional changes.

Monte Carlo simulation of additional safety control rod for commercial MNSR to enhance safety

A Monte Carlo simulation of additional safety rods for NIRR-1 HEU and LEU cores was carried out using the MCNP5 version 1.6 code. Two additional safety control rods having the same material composition as the main central control rod except for the surface area were studied. The following reactor core physics parameters were determined; neutron flux distribution within the core with safety rods withdrawn, control rod (CR) worth for each rod, core excess reactivity, shutdown margin and some kinetic parameters. Results obtained indicate that it would be feasible to include two additional safety control rods to improve safety level of the MNSR with little or no modification to the existing core configuration.

Design of an additional safety rod for Ghana Research Reactor-1 using MCNP5 code

Abstract In many licensed reactors, there are both control rods and shutdown or safety rods with the latter specifically designed for rapid shutdown of the chain reaction. But Ghana's Research Reactor 1 (GHARR-1), which is a Miniature Neutron Source Reactor (MNSR) has one control rod which serves the purpose of both reactivity control and shutdown of the reactor. This work presents an investigative study into the design of an additional safety rod, which will be an important safety component in the MNSR and will provide another alternative way to emergently shut down the reactor. The Monte Carlo Neutron Particle code; version 5, (MCNP5) was used to design the additional safety rod at one of the inner irradiation sites with cadmium as absorber material and stainless steel as clad. The irradiation site housing the safety rod was then enlarged, which affected the criticality of the core and a larger cadmium safety rod was inserted. In another instance, boron carbide was used as the absorber material also with stainless steel clad but without the further increment. After simulation, the cadmium safety rod, worth 2.94 mk could not shut down the reactor on insertion. However, the larger cadmium safety rod and the boron carbide safety rod of worth 4.54 mk and 5.90 mk respectively, attained subcriticality on insertion. The results indicate that an additional safety rod with boron carbide as absorber material can be used to enhance the safety of MNSRs.

The neutronic design of a critical lead reflected zero-power reference core for on-line subcriticality measurements in Accelerator Driven Systems

In the framework of the GUINEVERE project (Generation of Uninterrupted Intense NEutrons at the lead VEnus REactor) carried out at SCK⋅CEN in Mol, the continuous deuteron accelerator GENEPI-3C is coupled to the VENUS-F fast simulated lead-cooled reactor. With this installation the neutronic behaviour of an Accelerator Driven System (ADS) during different phases of operation is studied. Because of the subcritical character of ADS, an on-line reactivity monitoring system is required. The main goal of the GUINEVERE project is to design such a system, to evaluate its performance and accuracy, and to evaluate an extrapolation towards full-size ADS. The on-line subcriticality monitoring procedure within the GUINEVERE project requires a well-known critical reference state in order to determine the absolute subcriticality level of ADS. Direct absolute subcriticality measurement techniques can then be evaluated. The neutronic design of different types of critical configurations within the VENUS vessel is motivated, given the core fuel and lead reflector as basic elements. The neutronic analysis comprises operational aspects such as kinetic parameters, the weight of control and safety rods, the weight of a peripheral assembly, which are investigated by means of both probabilistic and deterministic neutronic codes. The outcome of this analysis determines the most suitable VENUS-F critical reference core. Critical reference configurations inside the VENUS vessel can be set-up with the available fuel (96 fuel assemblies) taking into account the uncertainty on the nuclear data. The flexible neutronic design allows an easy operation of the core. A sensitivity analysis on the nuclear data shows that an uncertainty corresponding to the weight of several peripheral fuel assemblies should be taken into account while designing this type of small cores.

Design, construction and simulation of a multipurpose system for precision movement of control rods in nuclear reactors

This article presents the design and implementation of a microcontroller-based system for the automatic movement of control rods in nuclear reactors of either power or research types. This system is controlled automatically, is linked to a personal computer system, and has manual controlling ability as well. The important features of this system are: automatic scram of the control rods, activation of alarm in emergency situations, and the ability to tune the control rod movement course both upwards and downwards. In this system, a small tank has been improvised as a coolant reservoir for pool type reactors such as Tehran Research Reactor and its water level is continuously adjusted by special sensors. Also, this system can be applied for controlling various types of control rods such as the regulating rods, safety rods and shim rods; can be connected to all reactor measurement tools and systems such as the period meter, power meter and flux meter; and can receive feedback signals from them. The devised system can be calibrated with these measurement tools by two special potentiometers in the related electronic board. The processes of this system have been simulated by the SIMULINK tool kit of MATLAB software and all responses of the system, including oscillation and transient responses, have been analyzed.

Design, Development, Testing and Qualification of Diverse Safety Rod and Its Drive Mechanism for a Prototype Fast Breeder Reactor

Prototype fast breeder reactor is U–PuO2 fueled sodium cooled pool type fast reactor and it is currently under construction at Kalpakkam, India. Prototype fast breeder reactor is equipped with two independent fast acting and diverse shutdown systems. A shutdown system comprises of sensors, logic circuits, drive mechanisms, and neutron absorbing rods. The two shutdown systems of prototype fast breeder reactor are capable of bringing down the reactor to cold shutdown state independent of the other. The absorber rods of the second shutdown system of prototype fast breeder reactor are called as diverse safety rods (DSRs) and their drive mechanisms are called as diverse safety rod drive mechanisms (DSRDMs). DSRs are normally parked above active core by DSRDMs. On receiving scram signal, the electromagnet of DSRDM is de-energized and it facilitates fast shutdown of the reactor by dropping the DSR into the active core. For the development of prototypes of DSR and DSRDM, three phases of testing, namely, individual component testing, integrated functional testing in room temperature, and endurance testing at high temperature sodium, were done. The electromagnet of DSRDM has been separately tested at room temperature, in furnace, and in sodium. Specimens simulating the contact conditions between electromagnet and armature of DSR have been tested to rule out self-welding possibility. The prototype of DSR has been tested in flowing water to determine the pressure drop and drop time. The functional testing of the integrated prototype DSRDM and DSR in aligned and misaligned conditions in air/water has been completed. The performance testing of the integrated system in sodium has been done in three campaigns. During the third campaign of sodium testing, the performance of the system has been verified with 30 mm misalignment at various temperatures. The third campaign has qualified the system for 10 years of operation in reactor. This paper presents design, development, testing, and qualification of the prototype DSR and DSRDM. Salient design specifications for both DSRDM and DSR are listed initially. The conceptual and detailed design features are explained with the help of figures. Details on material of construction are given at appropriate places. Test plans and criteria for endurance testing in sodium for qualification of DSRDM and DSR for operation in reactor are briefed. Brief explanation of test setups and typical test results are also given.

Testing and qualification of Control &amp; Safety Rod and its drive mechanism of Fast Breeder Reactor

Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control & Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements. CSR & CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot & dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism. In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM & CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM & CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO2 process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.