Reactor Power Change Research Articles

KATANA - water activation facility at JSI TRIGA, Part II: First experiments

Water as a primary coolant will play an important role in the performance of fusion reactors, as after being irradiated and activated, it causes an ionising radiation field throughout the facility. In direct support of ITER, the KATANA irradiation facility, which utilises a closed-water activation loop and serves as a well-defined and stable high-energy γ and neutron source, was commissioned in 2024 at the TRIGA Mark II research reactor at the Jožef Stefan Institute in Slovenia.A series of first experiments using the KATANA irradiation facility were performed to determine the operational characteristics of the KATANA, i.e. characterisation of the neutron flux within the irradiation component, dose rate measurements, spectrum characterisation of the activated water, and response to a change in reactor power & flow rate. The KATANA facility demonstrated the desired operational characteristics in terms of high and stable water flow rates and high activity values of the observed isotopes 16N, 17N and 19O, which is essential for minimising the experimental uncertainties.The first experiments performed at KATANA provided in-depth knowledge into the operation and capabilities of the facility and essential data to serve as a basis for further, more detailed/benchmark experiments.

Neutronics–Thermal-Hydraulics–Coupled Transient Analysis for Reactor Power Change in an Inclined Offshore Floating Boiling Water Reactor

An application of the boiling water reactor (BWR) to an offshore floating nuclear power plant (OFNP) in Japan is discussed. The BWR-type OFNP has some challenges for practical use, although it has high economic efficiency because of downsizing and simplification. One challenge is understanding reactor kinetics under conditions specific to the marine environment. This study quantitatively clarifies the total and spatial changes in power when the BWR is inclined during regular operation. Therefore, the TRACE (TRAC/RELAP Advanced Computational Engine) and PARCS (Purdue Advanced Reactor Core Simulator) codes were used to perform a three-dimensional neutronics–thermal-hydraulics–coupled transient analysis. The calculation model is based on Peach Bottom II. This study clarifies the changing trend in total and local BWR power by inclination with simplified modeling and conditions. The reasons for such changes are discussed based on changes in several thermal-hydraulic parameters. The difference in BWR power against the inclinations is small. Thus, it is implied that the BWR-type OFNP is expected to have a stable power supply capability during natural disasters. However, to confirm the power stability of the BWR reactor under a full range of offshore external conditions, further research is required. A description of additional research needs that would further support the safety case for this reactor design are discussed.

Transient behavior of a molten salt fast reactor under two-phase flow conditions with helium bubbling

In the present study, transient behaviors of a chloride molten salt fast reactor without control rods under two-phase flow conditions are analyzed assuming a bypass of the safety protection system. The two-phase flow in the core is caused by helium bubbles injected to remove the gases of the fission product. The void fraction under the rated condition is assumed to be as low as 0.2 %. However, as the fuel flow rate decreases, the void fraction increases inversely proportional to the flow rate and must be considered in any transient analysis. Therefore, analyses with a point reactor kinetics and thermal-hydraulics coupling model have been performed to account for void flow using the CFD code. Comparison between the transient under two-phase flow and the transient under single-phase flow in which the effect of reactivity due to voids is taken into account with the CFD analysis, shows almost the same results, although there are slight differences due to the difference in fuel volume corresponding to the void fraction. When analyzing transients with a system code that has difficulty handling two-component two-phase flow, the void effect has been considered using the above-mentioned single-phase flow analysis. In this way, the entire plant is analyzed using the RELAP5-3D code and the core inlet conditions are passed to the FLUENT code to analyze the transient behavior. The transients analyzed are UTOP, USBO, ULOFF, and ULOHS under the assumption of bypassing the safety protection system. In addition, a transient assuming the loss of helium bubble supply under the decay heat power conditions is analyzed and discussed. The results of the analysis are compared with the results of the same event without voids analyzed using RELAP5-3D. In the case of single-phase flow, the reactor power changes mainly due to temperature feedback. Although the two types of transients have similar appearance, the breakdown of reactivity is quite different. Relating the comparison between the calculated result using the system code and CFD code shows that for the short time behavior until the fuel flow rate shifts to low flow rate, the results of the CFD code, which can analyze the behavior of the helium voids, are more appropriate. However, there are no major problems in the same transient calculation using RELAP5-3D, which calculates the entire plant system.

Analysis of fluctuations in ex-core neutron detector signal in Krško NPP during an earthquake

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%–8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ∼4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ∼6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

Development of a coupled neutronics-CFD system for 540 MWE Indian Pressurized heavy water reactors (IPHWRs) and verification with slow online refuelling transient

A loosely coupled reactor is subjected to local perturbations due to reactivity device movement that affects the core neutronics in a local region of the reactor core. It is a challenge to predict the fuel-coolant characteristics like power, enthalpy, temperature etc., from a purely neutronics point of view hence a coupled reactor physics-CFD code is a must for the reactor. Additionally, the continual change in core composition due to on-power refueling along with coolant feedback throughout change in reactor power and/or reactivity device movement makes it clear that development of a coupled code is not sufficient in itself. The coupled code should capture the burnup effect of the fuel and the change in fuel heat transfer characteristics due to fission gas accumulation in the plenum gap. In this work a cross-section library with small step cross-section data for fuel and coolant has been developed. Simple interpolation gives a reliable data for neutronics computations. The coupled code discussed in this paper is an attempt to capture fuel characteristics accurately using a small step cross-section data and explicit computation of the effect of plenum gas in the fuel. The neutronics calculation is carried out in modified TRIVENI code and the CFD computation is carried out in Openfoam. Recent trends in reactor physics is towards high fidelity computing using Monte Carlo for pin-resolved neutron transport coupled with thermalhydraulics code. However, these high fidelity codes are computer resource intensive where as our code can be run on meagre computing resources. There is scope for improvement in not so well understood phenomena like plenum gas behaviour on the fuel characteristics and this work is highlighting this aspect of PHWR fuel.

АВТОМАТИЧНА СИСТЕМА КЕРУВАННЯ ЗМІНИ ПОТУЖНОСТІ ЯДЕРНОЇ ЕНЕРГЕТИЧНОЇ УСТАНОВКИ

Abstract. A special feature of nuclear energy production is the time of the process of controlled change in reactor power. Ensuring the stability of the nuclear power unit core when its power changes is a complex problem that can be solved by achieving shifts in the permitted range of AO in time. This allows to change the reactor power with maximum stability of the energy release in the core and reduces the error in modeling the static and dynamic properties of the nuclear power unit. The proposed automatic control system for changing the power of a nuclear power unit with a pressurized water reactor makes it possible to prevent fluctuations in the technological parameters of the reactor when switching control programs. Keywords: Automatic control system, axial offset, reactor power, nuclear power unit, pressurized water reactor.

Research on the effect of heat pipe temperature oscillation on heat pipe cooled reactor operation

Heat pipes have been recognized as the unique equipment of heat absorption for heat pipe cooled reactor. The operating characteristics of heat pipe have the direct effect on the operation of the reactor system. In accordance with the results of the theoretical analysis and the experimental research, it has been confirmed that the heat pipe will have the temperature oscillation whether it has the horizontal operation or the inclined operation. In this study, the effect of temperature oscillation on heat pipe cooled reactor was preliminary investigated. The typical heat pipe cooled reactor named Kilowatt Reactor Using Stirling TechnologY (KRUSTY) was chosen for study. To simplify the analysis, the operating characteristics of heat pipe were approximated by setting the wall temperature change of condenser. As revealed by the simulated results, when the temperature of heat pipe oscillated, the fuel temperature would oscillate with the similar period leading to the significant periodic change of reactor power. Besides, the heat absorption of heat pipes would be significantly affected. Meanwhile, The longer the period and the larger the amplitude, the more obvious the effect on the reactor operation. For the KRUSTY reactor, even if the fuel temperature oscillation was only 8.0 K, there would be nearly 560 W change of reactor power which is about 12 % of the fission power. Moreover, the mitigation effect using the buffer layer was discussed. The simulated results suggested that the buffer layer with a smaller thermal diffusion coefficient could effectively reduce the temperature oscillation and the power oscillation, whereas it would lead to the higher reactor operating temperature, which might reduce the safety of the reactor system. All the discussions and analyses could contribute to the design of the heat pipe cooled reactor system and to the understanding of the real operating characteristics of this reactor system.

Possible Implementation of Ex-Core Measurement in TEPLATOR Graphite Reflector

Abstract An ex-core neutron flux measurement is a crucial system for all common power reactors. It is necessary to monitor the neutron flux and control the chain reaction, therefore, the ex-core neutron flux measurement is one of the main safety and control systems. The main advantage of this arrangement of detectors is a fast response to neutron flux change, which determines the reactor power change. Regarding to the new reactor concepts, it is important to deal with improved detection systems suitable for these reactors. Many of the modern reactor concepts are based on a graphite moderator or reflector, which is also the case of the TEPLATOR. The TEPLATOR is a solution of a district heating system based on heavy water as a moderator and graphite as a reflector. The TEPLATOR is designed to use irradiated fuel from the commercial pressurized water reactor or boiling water reactor, which has low to intermediate burnup. This work is focused on the verification of the possible use of the special neutron measuring system placed in the graphite reflector. The Monte Carlo code serpent was used for the calculations performed in this work.

Modeling of neutronics and thermal-hydraulic aspects during total instantaneous blockage of central fuel subassembly in a fast reactor

Studies on local flow blockages are of great importance in Sodium cooled Fast Reactors (SFRs). Total instantaneous blockage (TIB) at the inlet of a subassembly (SA) of an SFR is a beyond design basis event that is considered as a theoretical envelope of all small blockages. The extent of core damage propagation is an essential input for the thermal design of the core catcher, which is provided for the collection and cooling of molten core. This study aims to conduct a coupled thermal hydraulics and neutronics analysis of TIB in a medium sized SFR up to the end of fuel melting in the blocked SA. Towards this, a coupled thermal hydraulics and neutronics model is developed by solving heat transfer equations and point kinetics equations numerically. A one-dimensional (1-D) molten clad film motion model is incorporated for simulating the clad relocation dynamics. Using the developed models, a coupled analysis of TIB in a medium sized SFR is carried out. The time of occurrence of key events such as sodium boiling, clad melting, and fuel melting are predicted. The times of detection of the accident by the core temperature monitoring system of neighbouring SA and by the overpower detection signals are also predicted. The time of the TIB detection is earlier than that predicted by a previous study which does not consider reactor power change during the transient. The rise in the sodium outlet temperature beyond the threshold level for the TIB detection is observed for all the fuel SAs before the end of fuel melting in the blocked SA. Thus, assurance towards early detection of TIB by core temperature monitoring system is strengthened.

A comparative study of reactivity feedbacks including in-pin motion and their impact on UTOPA in a 600 MW fast reactor

The design experience of PFBR (Prototype Fast Breeder Reactor) gives confidence in designing advanced Sodium cooled Fast Reactors (SFR) with improved economy and passive safety features towards enhanced safety. In this regard, Fast Breeder Reactor (FBR) is optimised with 600 MWe leading to better economy. To prevent severe core damage & large radioactivity release to the public and also to enhance the safety, improved inherent safety characteristics are adopted with passive shutdown features. Important safety measures taken to improve inherent safety are the optimization of sodium void reactivity beneath 1 $, which is in line with the current international thinking. To incorporate the enveloping beyond design basis events (BDBE) within the Design Extension Condition (DEC), additional passive safety mechanisms are envisaged to strengthen the reactor safety and practically eliminate the probability of Core Disruptive Accidents (CDA). One such BDBE is Unprotected Transient Over Power Accident (UTOPA), in which uncontrolled withdrawal of one control rod leads to rise in power and fuel temperature. Under UTOPA, it is possible for fuel to melt and undergo in-pin fuel motion through the available central hole. Upon initiation of melting, fission gases trapped within the fuel microstructure are released into the pellet cavity along with molten fuel. Consequently, a multi-phase flow occurs inside the pellet cavity, with several hydrodynamic effects influencing the fluid motion. The relocation of molten fuel from high fuel void worth to low fuel void worth region results in a negative reactivity feedback. This in-pin fuel motion or squirting feedback reduces the total reactor power as well as the hot-spot clad and coolant temperatures. With change in reactor power during UTOPA, the balance of the plant gets affected. This results in a change in the inlet coolant temperature, which affects the overall core temperature profile and its respective feedbacks. Considering the above said arguments, UTOPA study has been carried out with different initial conditions such as Constant Inlet Coolant Temperature (CICT), Time Dependent Inlet Coolant Temperature (TDICT), with and without In-Pin Fuel Motion (IPFM) feedback. Comparisons of the results are made to get better understanding of various feedbacks and their impact on UTOPA, highlighting the importance of in-pin fuel motion reactivity feedback in a medium sized fast reactor core. From the study, it is concluded that the case without considering IPFM feedback gives the most conservative results and if the transients converge to a safe state without considering IPFM feedback then, the transients are expected to settle down to a safe state with IPFM feedback.

Experimental analysis of step reactivity insertion effect on reactor power, fuel temperature and reactor period in BAEC TRIGA research reactor

A series of experiments has been performed in a 3 MW BAEC TRIGA Mark-II Research Reactor to analyze the effects of positive and negative step reactivity insertion. The responses of reactor power and fuel temperature were determined. The reactor period is also calculated and a comparison is shown how the initial power affects the reactor period. This research illustrates the behavior of the delayed neutron in response to step reactivity insertion. After sudden step reactivity insertion (positive or negative), a sharp change in reactor power and fuel temperature has occurred initially, then the variation slows down followed by stable reactor period. In spite of inserting same positive and negative step reactivity, the amount of subsequent increase and decrease in reactor power and fuel temperature were not same. Significant variations in the reactor power and reactor period are found due to the effect of low initial power.

Development of Device for Measuring the Heat Generation Rate of Materials in Research Reactor

To explore the axial distribution of the heat generation rate of the materials in the channels of the test reactor, a heat generation rate measuring device was designed, by taking the G7 channel of the high flux engineering test reactor (HFETR) as an example. The strain distribution contours of the device and test section under load are obtained by numerical simulation method, and the temperatures of the calorimeter to calibrate the bridge and the measurement bridge are obtained by physical calculation. The heat generation rate measurement is carried out in the G7 channel using this test device. The results show that the overall structure of the device meets the strength requirements, and a protective tube needs to be installed between the calorimeters in the test section. The calculated temperature of the sample and the bridge is lower than the melting point of the material, and the device meets the thermal requirements. The heat generation rate measured by the test varies with the reactor power changes regularly. Different materials have different γ absorption properties under different energy levels of γ-ray environment. The device can be used as a heat generation rate measurement tool to provide a guarantee for determining the heat generation rate distribution of different materials in the nuclear reactor.

Reactor dynamic simulation to analyze possible scenarios after a spurious opening of the safety flapper valve of TRR during the normal operation regarding the inherently safety features and Design Extension Conditions (DEC) by using the RELAP5 code

LOFA is one of the most important PIE in nuclear reactors. For the pool-type research reactor, the safety flapper valve is usually located below the reactor core and have a very important rule on the reactor safety especially during the loss of the pump accident. In this paper, TRR has been simulated and analyzed against the spurious opening of the safety flapper valve during normal full power operation. First of all, the steady state parameters are successfully benchmarked against the operational data of TRR. Then, the reactor dynamic calculations are successfully benchmarked against a sophisticated experimental dynamic test to evaluate the reactor power changes due to reactivity feedback. Then three different types of possible scenario including both types of DBA and DEC are simulated and analyzed. Furthermore, inherently safety features are evaluated against temperature rises. Results indicate that if the anticipated emergency shutdown signal is triggered successfully, the reactor core remains safe against any physical damage. If the reactor reactivity control system is completely disabled, the inherently safety features of the reactor decrease the reactor power effectively via negative reactivity feedback and the reactor fuel assemblies remain safe again. But if the fault diagnosis system of the spurious opening of the safety flapper valve fails to detect the occurrence or it does not send the emergency shutdown signal, and at the same time, the automatic power regulation system keeps the normal operating conditions, the hot spot may reach to the fuel melting point just a few seconds after the accident occurrence.

Assessment of neutronic safety parameters of VVER-1000 core under accident conditions

The major issue in designing safety and control systems for power reactors is preventing accidents or reducing the hazard imposed. Any change in reactor power causes a change in temperature and material densities of the system that will affect the reactivity, which in turn, causes transients/accident conditions. The analysis of main safety parameters is essential for any change to ensure the safe operation of the reactor in both steady and transient situations. In this paper, the impact of temperature, moderator density, boron concentration and control rods position changes on the safety parameters is studied to assess the behavior of the reactor under accident conditions. The main safety parameters as temperature reactivity coefficients, power peaking factor, effective delayed neutron fraction and generation time are evaluated under accident conditions for VVER-1000 core using MCNP6 code and ENDF/B-VII.1library. The results evidenced that the reactivity coefficients are still negative, under accident conditions, which guarantee the safe operation of the reactor. The effective delayed neutron fraction increased as a result of increased temperature, decreased moderator density and addition of the soluble boron. The worth of the control rods is sufficient to make the reactor into a subcritical state either in emergency situations or with regular shutdown.

Computational burnup analysis of the TRIGA Mark II research reactor fuel

In this study, analysis of the complete operational history of the “Jožef Stefan” Institute (JSI) TRIGA reactor was performed. Reactor power changes, core configurations and weekly excess reactivity measurements were analysed to obtain the needed data for fuel burnup calculations. More than 50 years of reactor operation was simulated using deterministic code TRIGLAV and stochastic code Serpent-2. The calculated core reactivities are in good agreement compared with the excess reactivity measurements. Code-to-code comparison is presented. Clear agreement is observed when comparing changes in core excess reactivity, and discrepancies are observed in the comparison of individual fuel element burnup and its isotopic composition. The Serpent-2 results are in better agreement with the measurements compared to the TRIGLAV code; nevertheless, a conclusion can be made that the TRIGLAV code is viable for TRIGA fuel management and burnup analysis. A three-dimensional (3D) burnup study was conducted, where individual fuel elements were further divided into multiple angular and axial depletion zones. Notable burnup effects were observed, and an explanation using surrounding water distance is presented.

Design of a FPGA-based controller for power and period measurement in the start range of Dalat Nuclear Research Reactor

This paper introduces a new controller module based on a high-speed field-programmable gate array (FPGA) and digital signal processing (DSP) using moving average (MA) filters for calculation of the reactor power and period at the start range of the Dalat nuclear research reactor (DNRR). The reactor power is proportional to the neutron flux in the reactor core, and the reactor period is the time that the reactor power changes by a factor of 2.718. In the control and protection system (CPS) of the DNRR, the reactor power and period have been monitored by the 8-bit microprocessor controller named BPM-107R. There are two main functions of the BPM-107R controller including 1) measurement and determination of reactor power and period and 2) generation of warning and emergency protection signals by reactor power or/and by reactor period. Those discrete signals will access to the logical processing unit of the CPS to prohibit the upward movement of control rods or to shut down the reactor. The CPS has three BPM-107R controllers corresponding to three independent neutron flux measurement equipment (NFME) channels working by logic voting “2 out-of 3”. Each NFME channel was designed for detection of neutron flux density in the full range from 1×100 to 1.2×1010 n/cm2 ×s, which is divided into two sub-ranges named start range (SR) and working range (WR). The designed FPGA-based controller module was tested using simulated signals as well as signals from the CPS in comparison with the original controller BPM-107R. The experimental results show that the characteristics and functions of the two controllers are equivalent.

Research of WWER-1000 Power Change Modes for Operation in the Load-Following Mode

Ukrainian nuclear power plants (NPPs) generate approximately 55 % of all the country’s electricity production in the basic mode. The Ukrainian energy system faces power shortage for load-following mode (LFM). Hence, operation of Ukrainian NPPs in LFM is advisable. The second stage of LFM pilot operation has been completed at Khmelnitsky NPP Unit 2 to date.
 Following the above, the research of LFM implementation key aspects at Ukrainian NPPs is urgent and relevant. One of the key aspects is the definition of the appropriate mode for reactor power change in LFM for common type of the Ukrainian reactors (WWER-1000/V-320).
 Power density field and thermal power in WWER-1000 are controlled by means of boron control and control rods. The paper provides investigation of the standard power change modes and power density field control in accordance with the safety requirements in the context of the boric acid volume used to decrease the power level, the axial offset stability, maximum values of the power peaking factor and fuel utilization performance (parameters).
 The computer data of the parameters are obtained in SE «NNEGC «Energoatom» during neutron-physical calculations of WWER-1000 cores by means of the «BIPR-7A» program, that uses a mathematical model of WWER reactors. The calculation was conducted for three modes:
 Mode 1: CRs of group 10 and boron control. The mode is being operated for LFM in Ukraine. Group 10 and boric acid/distillate water (BC) are used for power change and power density field control.
 Mode 2: Only CRs of group 10. Only group 10 is used for power change and power density field control.
 Mode 3: Only CRs of groups 10, 9 and 8. All the A-algorithm control groups are used for power change and power density field control.
 Compensation of the slow reactivity variations due to the xenon poisoning and fuel burn up before/after power change was provided by boric acid for all modes.
 The research define the appropriate mode in the context of the parameters.
 The research determine the advantages and drawbacks of the considered modes in the context of the parameters, as well as ways to improve the power density field control and power change system of WWER-1000 core.

Impact of nuclear reactor radiation on the performance of AlN/sapphire surface acoustic wave devices

The performance of an AlN/sapphire surface acoustic wave (SAW) delay line device was characterized in real time under irradiation inside a nuclear reactor. Both its resonant frequency and transmission efficiency were observed to respond to a change in reactor power. The response follows an exponentially saturating behavior after a step power increase, followed by an exponentially decaying recovery after reactor shutdown. A sensitivity analysis based on the governing electro-mechanical equations shows that the frequency shift can be attributed to the softening of sapphire’s elastic constants under neutron radiation. A kinetic rate equation is adopted to interpret device response and describe its microstructural evolution. These results suggest that the AlN/sapphire SAW device remains functional under irradiation, is sensitive to neutron and gamma ray fluxes, and offers an opportunity for remote sensing and in-situ measurement of material properties when exposed to nuclear reactor environment.

Participation of Ukrainian NPPs in Regulating Frequency and Power in the United Energy System: Problem Analysis and Solutions

The issue related to the participation of Ukrainian nuclear power plants in regulating frequency and power in the united energy system of Ukraine is considered in the paper. The dependence of frequency changes in the grid on the imbalance between generation and consumption of active power is presented. This can violate conditions of parallel operation of generators. Three levels of the frequency control system in the power system using backup capacities are examined. The issues of participation of nuclear power units in regulating frequency and power are studied. It is shown that the main factors limiting such participation are the characteristics of fuel assemblies. These characteristics govern the change in reactor power in a narrow range. An obstacle is also the mismatch of the characteristics of the equipment of power units for participation in regulating power system frequency. The complexity of regulating the electric power of a WWER nuclear power unit is under consideration, which is implemented by coordinated regulation of the reactor power and generator power. The regulatory requirements for the characteristics of the primary frequency regulation by nuclear power units is carried out. The possibility of their participation in the normalized primary frequency regulation are analyzed. The paper considers the concept of the National Nuclear Energy Generating Company “Energoatom” for the modernization of systems and equipment of WWER-1000 power units, which provides bringing the characteristics of power units in line with standards. The methodology for preliminary testing of the readiness of Zaporizhzhya NPP Unit 1 (“pilot” power unit) to participate in such regulation is described. The methodology involves testing at two power levels at the beginning and in the end of the fuel campaign. A signal is used to simulate a frequency deviation that is input via a separate train of the control system. The tasks, conditions and criteria for the success of the tests are described. Based on the test results, conclusions will be drawn about the possibility of the safe participation of WWER-1000 power units in the normalized frequency regulation in the united energy system of Ukraine and the necessary measures to modernize the equipment.

Gamma heat analysis in various power levels of RSG G.A. Siwabessy silicide core

Reaktor Serba Guna G.A. Siwabessy (RSG-GAS, previous name MPR-30) is the largest research reactor in Southeast Asia that acts as a national facility to irradiate material. Gamma heat is a very important factor for the safety analysis in every material irradiation activity at the irradiation facility. Gamma heat is the main research topic of several forms of safety investigations at world research reactors. Gamma heat information is useful to predict the temperature of the material to be irradiated and negligence in predicting gamma heat can cause overheating. Gamma heat values are very dependent on the characteristics of the reactor core. Changes in reactor power can affect the core characteristics. RSG-GAS is designed to have a 30 MWth of nominal power but it is currently operated at 15 MWth power level. In this study observed changes in gamma heat as a function of reactor power and material target on the Central Irradiation Position (CIP) silicide core of RSG-GAS by using a modified GAMSET program. Modifications are made by adjusting the material and power configuration in the core. The results of the analysis show that gamma heat will be increase in accordance with the increase in power level in various material targets such as graphite (C), aluminium (Al), iron (Fe) and zirconium (Zr). Gamma heat also tends to increase according to the increase of atomic number target. Verification has been carried out with the results of calculations in the 35 MWth CEA Grenoble reactor with the smallest yield difference of 1.23% in the graphite target and 2.71% in the iron target.