Reactor Scram Research Articles

Space–time convergence analysis on BWR stability using TRACE/PARCS

Unstable behavior of Boiling Water Reactors (BWRs) is known to occur during operation at certain power and flow conditions. Even though BWR instability is not a severe safety concern, it could cause reactor scram and significantly decrease the economic performance of the plant. This paper aims to (a) quantify TRACE/PARCS space–time discretization error for simulation of BWR stability, (b) establish space (nodalization) and time discretization necessary for space–time converged model and (c) show that the space–time converged model gives more reliable results for both stable and unstable reactor. The space–time converged model is obtained when further refinement of numerical discretization parameters (nodalization and time step) has negligible effect on the solution. The study is significant because performing a space–time convergence analysis is a necessary step of qualification of the TRACE/PARCS model, and use of the space–time converged model increases confidence in the prediction of BWR stability.

Development of three-dimensional hot pool model in a system analysis code for pool-type FBR

A three-dimensional hot pool analysis code with SIMPLE algorithm was developed based on staggered grid, which can predict thermal-hydraulic characteristic in hot pool of pool-type liquid metal fast breeder reactor (LMFBR) under Cartesian and cylindrical coordinate systems. After being incorporated into system analysis code for pool-type fast reactor in China (SAC-CFR), the coupled code was used to analyze the thermal-hydraulic behavior in the upper plenum of fast breeder reactor “MONJU” during reactor scram transient. A basic agreement was obtained, which means the present model is effective. And then the coupled code with newly developed porous medium model was used to analyze the flow field in China Experimental Fast Reactor hot pool under steady-state operation condition. The distribution characteristic of flow field in hot pool showed the effectiveness of porous medium model, which formed preparations for further development of passive residual heat removal system in-vessel.

Analysis of BR2 Loss-of-Flow Tests C and F

To support the safety analysis of the conversion of the BR2 research reactor to low-enriched uranium (LEU) fuel and extend the validation basis of the RELAP code for the analysis of the conversion of research reactors from highly enriched (HEU) fuel to LEU, the simulation of BR2 tests A/400/1, C/600/3, and F/400/1 was undertaken. These tests are characterized by loss of flow initiated at different reactor power levels with or without loss of system pressure, reactor scram, flow reversal, and reactor cooling by natural circulation. This work presents the RELAP analysis of tests C/600/3 and F/400/1 and comparison of code predictions with experimental measurements for peak cladding temperatures during the transient at different axial locations in an instrumented fuel assembly. The simulations show that accurate representation of the power distribution, especially after reactor scram, between the fuel assemblies and the moderator/reflector regions is critical for the correct prediction of the peak cladding temperatures during the transient. Detailed MCNP and ORIGEN simulations were performed to compute the power distribution between the fuel assemblies and the moderator/reflector regions. With these distributions the predicted peak cladding temperatures are in good agreement with experimental measurements.

Calculation of reactivities using ionization chamber currents with different sets of kinetic parameters for reduced scram system efficiency in the VVER-1000 of the third unit of the Kalinin nuclear power plant at the stage of physical start-up

The effectiveness of the VVER-1000 reactor scram system is analyzed using ionization chamber currents with different sets of kinetic parameters with allowance for the isotopic composition in the calculation of these parameters. The most “correct, aesthetically acceptable” results are obtained using the eight-group constants of the ROSFOND (BNAB-RF) library. The difference between the maximum and minimum values of the scram system effectiveness calculated with different sets of kinetic parameters slightly exceeds 2β. The problems of introducing corrections due to spatial effects are not considered in this study.

Analysis of the BR2 Loss-of-Flow Test A

The simulation of the BR2 test A/400/1 was undertaken to support the safety analysis of the conversion of the BR2 research reactor to low-enriched uranium (LEU) fuel and to extend the validation basis of the RELAP code for analysis of the conversion of research reactors from highly enriched fuel to LEU. This test was characterized by a steady-state peak heat flux of 400 W/cm2, total loss of flow without loss of system pressure, reactor scram, flow reversal, and reactor cooling by natural convection. This paper presents the RELAP analysis of test A/400/1 and the comparison of code predictions with experimental measurements of peak cladding temperatures during the transient at different axial locations in an instrumented fuel assembly. The simulations show that accurate representation of the pump coastdown characteristics and of the power distribution, especially after reactor scram, between the fuel assemblies and the moderator/reflector regions are critical for correct prediction of the peak cladding temperatures during the transient. Detailed MCNP and ORIGEN simulations were performed to compute the power distribution between the fuel assemblies and the moderator/reflector regions. With these distributions, the predicted peak cladding temperatures were in a good agreement with experimental measurements.

Control strategies for transients of hydrogen production plant in VHTR cogeneration systems

Operability of Very High Temperature Reactor (VHTR) hydrogen cogeneration systems in response to abnormal transients initiated by the hydrogen production plant is one of the important concerns from economical and safety points of views. The abnormal events in the hydrogen production plant could initiate load changes and induce temperature variations in a primary cooling system. Excessive temperature increase in the primary cooling system would cause reactor scrams since the temperature increase in the primary cooling system is restricted in order to prevent undue thermal stresses from reactor structures. Also, temperature decrease has a potential propagation path for reactor scrams by reactivity insertions as a consequence of the reactivity feedbacks. Since suspensions of reactor operation and electricity generation should be avoided even in case of abnormal events in the hydrogen production plant from an economical point of view, an establishment of a control scheme against abnormal transients of hydrogen production plant is required for plant system design. In the present study, basic controls and their integration for the GTHTR300C, a VHTR cogeneration system designed by JAEA with a direct Brayton cycle power conversion unit and thermochemical Iodine–Sulfur process hydrogen production plant (IS hydrogen production plant), against abnormal transients of IS hydrogen production plant are presented. Transient simulations for selected load change events in the IS hydrogen production plants are performed by an original system analysis code which enables to evaluate major phenomena assumed in process heat exchangers of the IS hydrogen production plant. It is shown that abnormal load change events are successfully simulated by the system analysis code developed. The results demonstrated the technical feasibility of proposed controls for continuous operation of the reactor and power conversion unit against load change events in the IS hydrogen production plant.

Integrated control system responses during load rejection transient for Lungmen ABWR Plant

The RETRAN model of Lungmen ABWR Plant was used to evaluate the load rejection with bypass transient of the start-up test program. There are three major control systems implemented in Lungmen RETRAN Model, including recirculation flow control, feedwater control, steam bypass and pressure control system. The responses of the control systems are important to help the reactor operate smoothly in the test, and credited to mitigate the transient impact. A brief description about the Lungmen RETRAN Model and the design feature of control systems is included in this article. The predicted analysis of load rejection test was conducted by using RETRAN-3D code to verify the dynamic responses of the reactor and control systems to protective trips in the generator, and also to demonstrate the capability of the plant to withstand against a generator load rejection without reactor scram. The analysis results show that the reactor has not been scrammed due to the proper control system responses, and sustained stable operation under the condition of the minimum station house load. The sensitivity studies show that there will be no significant pressurization in load rejection transient as long as activating bypass capacity functions in time. While, the sensitivity study cases of other settings in the mitigation facilities have no outstanding impact on the system response.

ASTEC V2.0 computational evaluation of source term and its chemical forms under accidental conditions during mid-loop operation

The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic safety studies as a major contributor to core melt frequency and source term risk. The fission products release and transport to the containment has been analyzed during mid-loop operation of a reference PWR 1000 MWe reactor using the severe accident integral code ASTEC V2.0. The analyses have been performed considering the loss of residual heat removal (RHR) system at various times after reactor shutdown for the reactor vessel configuration with the removed upper head (open reactor). In this configuration, the possible air ingress can have an impact on safety such as accelerated oxidation and increased volatility of certain FPs (particularly iodine and ruthenium). Sensitivity calculations have been performed in terms of air ingress simulation with a different intensity. Besides equilibrium chemistry model, most of the calculations have also used a limited kinetics model. The study has shown that without air ingress the only predicted gaseous form of iodine is HI (≤7.4% of the total mass of iodine released from core) and no gaseous RuO 4 is created. Sensitivity calculations have illustrated that the gross fraction of gaseous iodine (I 2 + HOI + HI) has an increased trend with growth of air ingress intensity and with the duration of sequence evolution. In most oxidative atmosphere the gross iodine gaseous fraction could increase by a factor form of two to several times as compared to the corresponding case without air ingress (particularly due to I 2 persistence). Creation of gaseous RuO 4 is sensitive to carrier gas temperature; therefore a considerable fraction (≤3%) is predicted only in the sensitivity cases with the shortest time of loss of RHR after reactor scram.

ICONE19-43754 RETRAN Analysis of Feedwater Controller Failure Transient for Lungmen ABWR

This study used RETRAN-3D program to analyze the FWCFWB transient at 100% rated power/111% rated core flow and under EOC operating condition where the ΔCPR response was worst. Important system operational actions included (1) the turbine and the feedwater pumps tripped on high water level (L8), (2) fast bypass valves opening which inhibited reactor scram and RIPs trip on turbine trip, (3) reactor scram and the four RIPs which are not connected to the Motor Generator (M-G) sets tripped on low water level (L3), (4) low-low water level (L2) initiation of the Reactor Core Isolation Cooling System (RCIC). The ΔCPR of this event was 0.05 and the results were compared with those from FSAR. It showed that there was a significant difference in the change rate of vessel water level, but other transient trends were in good agreement with FSAR. A test of all bypass valves failed to open was also covered in this study. In this situation, the reactor scram occurred on high water level turbine trip, and the ΔCPR would rise to 0.17.

Simulation of the EBR-II Loss-of-Flow Tests Using the MARS Code

As part of the development of a safety analysis methodology for a liquid-metal reactor (LMR) in Korea, the Multidimensional Analysis of Reactor Safety (MARS) code was selected as a system transient safety analysis code. The Korea Atomic Energy Research Institute developed the MARS code to analyze safety and thermal-hydraulic phenomena related to a two-phase flow in the transients of water reactors a decade ago. The addition of thermal-hydraulic models related to liquid metal as a coolant and reactivity feedback models associated with the kinetics calculation of an LMR core is required for the application of the MARS to the transients of an LMR design. A table for various properties of liquid sodium, several heat transfer coefficients according to flow regimes and geometries, and the models for a pressure drop due to the wire spacers of the LMR core were newly implemented. The improved MARS code was verified through the analysis of three shutdown heat removal tests (SHRT)-17, -39, and -45 conducted in the Experimental Breeder Reactor (EBR)-II reactor. The SHRT-17 test involved a simultaneous loss of electrical power to all pumps and a reactor scram from 100% power and flow. Thus, the test simulated a thermal-hydraulic transition from a forced convection to the totally passive decay heat removal due to a natural circulation. SHRT-39 and SHRT-45 are loss-flow tests without a reactor scram. However, the pump coastdown periods and initial states of the plant are different from each other. Simulated results for the flow rate and temperature for an instrumented subassembly agree well with the experimental data.

Estimation of SCRAM Rate Trends in Nuclear Power Plants Using Hierarchical Bayes Models

Nuclear reactors are equipped with reactor scram systems to ensure rapid shutdown of the system in the event of leaks, failure of power conversion systems, or other operational abnormalities. The U.S. Nuclear Regulatory Commission (NRC) collects data on scram rates from various nuclear power plants over time to estimate the trend of proper functioning of the plants which in turn is used to regulate them. The annual scram data obtained from 66 commercial nuclear power plants indicate an increase in the number of plants having no scrams from 1.5% in 1986 to 33% in 1993. To analyze correlated count data with excess zeros (e.g., no scrams), a zero-inflated model that accounts for both temporal and plant-to-plant variation is being developed in this article. A wide class of possibly non-nested models was fitted using Markov Chain Monte Carlo (MCMC) methods and compared using a predictive criterion. Out-of-sample tests were also performed to study the performance of the models in predicting the scram rates of the plants. For the NRC data on scram rates, the stochastic time trend models that account for zero-inflation were found to provide a much better fit compared to the deterministic trend models.

Post-test analysis of helium circulator trip without scram at 3 MW power level on the HTR-10

Safety demonstration tests were conducted on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the transient data of reactor core and primary cooling system for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 3 MW power level on October 15, 2003. This paper simulates and analyzes the power transient and the thermal response of the reactor during the test by using the THERMIX code. The analytical results are compared with the test data for validation of the code. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shut down after the stop of the helium circulator; the subsequent phenomena such as the recriticality and power oscillations are also studied. During the test a natural circulation loop of helium is established in the core and the other coolant channels and its consequent thermal response such as the temperature redistribution is investigated. In addition, temperatures of the measuring points in the reactor internals are calculated and compared with the measured values. Satisfactory agreements obtained from the comparison demonstrate the basic applicability and reasonability of the THERMIX code for simulating and analyzing the helium circulator trip ATWS test. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature during the test is always lower than 1600 °C which is the limited value for the HTGR.

Simulations and modeling of radioactivity buildup due to corrosion products in ion exchanger and filters of a typical PWR

A set of coupled first-order differential equations have been developed to simulate the corrosion product activity in ion exchanger and filters of a typical pressurized water reactor (PWR). Computer program CPAIR-RC has been modified for the current study to include the effect of nonlinear acceleration of corrosion on corrosion product activities. Simulations for five corrosion products 59Fe, 99Mo, 56Mn, 58Co, and 60Co have been carried out. Simulation results show that the activities due to 59Fe, 99Mo, and 56Mn saturate within approximately 1000 h of reactor operation. However, the activities due to 58Co and 60Co keep on accumulating in the ion exchanger of PWR during this time. The activity buildup in the ion exchanger due to 56Mn starts after 77 h of reactor operation. When flow rate perturbations are introduced by decreasing flow rate to 40% of rated value, the activity due to 56Mn saturates at the value of 1.27 × 10−3 μCi cm−3. For the case of nonlinearly acceelerating corrosion with a pump coast down, the activity for 59Fe shows an initial drop with a decrease in flow rates. The activity drops from 2.37 × 10−2 to 2.33 × 10−2 μCi cm−3 before the reactor scram. The corrosion product activities calculated by the CPAIR-RC program are compared with some of the experimental data available in literature and show fair agreement.

Simulation of LOFA and RIA for the IEA-R1 research reactor using the code MERSAT

The thermal hydraulic system code MERSAT has been applied to model the Brazilian IEA-R1 research reactor. The developed MERSAT model consists of detailed description for the primary loop including reactor core, bypass, reactor pool, main pump and coupling valve. The developed model is verified through the simulation of the steady state operation at the nominal power of 5 MW. According to the safety significance of postulated initiating events two selected design basis accidents corresponding to loss of flow accident (LOFA) with reactor scram and reactivity insertion accident (RIA) without scram have been analyzed. The results of LOFA simulation demonstrate the phenomena of flow reversal characterizing the flow turnaround from forced downward to natural upward circulation. The prevailing natural circulation after the flow reversal guarantees the removal of residual decay heat. The maximum clad temperature of the hot channel reaches its maximum of about 73 °C after 35 s after the initiation of the transient. The step reactivity insertion of 2 mk leads to increasing the reactor power up to 1.7 times the nominal value. Consequently, the maximum clad temperature of hot channel reaches about 99 °C which is still far below the onset of subcooled boiling. Hence, the established thermal hydraulic conditions ensure a considerable safety margin even if the reactor protection system should fail to shutdown the reactor 400 s after the begin of the reactivity insertion event. The results of LOFA and RIA demonstrate the safety features of IEA-R1 under the investigated conditions. In both cases the thermal hydraulic design limits have not been exceeded. However, for the RIA case the maximum clad temperature exceeds the limit of 95 °C being chosen as operation criteria for limiting corrosion on the surface of the fuel elements.

A GFR benchmark: Comparison of transient analysis codes based on the ETDR concept

In preparation for the transient analysis of the Generation IV gas fast reactor (GFR) and experimental technology demonstration reactor (ETDR) designs, a transient benchmark exercise (in the frame of the Generation IV GFR project) was performed to compare the capabilities and limitations of the different code systems to analyse these new reactor concepts. The benchmark was based on the ETDR concept and was performed in three phases, i.e. a loss-of-flow (LOF) transient with reactor scram for phases 1 and 2 and a small-break loss-of-coolant accident for phase 3. The organizations which participated in the benchmark were AREVA, France; CEA, France; NRG and TUD, The Netherlands; AMEC, United Kingdom; INL, USA (phase 1); CIRTEN, Italy; JRC-IE EURATOM and PSI, Switzerland. Phase 1 of the benchmark was performed in a “blind” manner in that all participants were provided with the same information, but were free to make their own judgement on the use of this information. Following a review of the results from phase 1, the conclusion was made that the agreement between all the partners was acceptable, but better agreement was to be preferred. The reason for the discrepancies was identified, namely: different heat transfer correlations for the main and DHR heat exchangers and reactor core, modelling of the vessel wall thermal capacity, core and vessel pressure drop calculations and flow resistance in the DHR helium and water loops. The phase 1 transient was repeated as (phase 2) but with more ‘strict boundary conditions’ to identify which of the above gave the largest contribution to the differences in the phase 1 submissions. The second phase results were much improved. However, to demonstrate the capabilities of the codes at low pressures a SB-LOCA transient with reactor scram (phase 3) was defined to model the ETDR DHR system during the depressurization from 70 bar to the 3 bar containment pressure. The phases 1 and 2 results showed that the use of different heat transfer correlations for the main heat exchanger does not significantly affect the core coolant temperatures, while the modelling of ETDR heat structures has a major impact on the core inlet temperatures. Even though three participants used the RELAP5 code and two participants used CATHARE, there was limited consistency between the results for each specific code. This was particularly true for phase 1, where the wide spread in the results came from different input and boundary assumptions by the codes users, i.e. the so-called “user-effect”. The results of the third phase (SB LOCA with reactor scram), showed sensitivity to the extrapolation of the core pressure drop calculation to low pressures and low Reynolds numbers, with the result that there was an extreme sensitivity of the core flow rate to small changes in core geometry, grid spacer losses, etc. resulting in large differences in coolant, clad and fuel temperatures. The most important message to be taken from the benchmark exercise is that to obtain consistent results, it is especially important to have consistency in the use and interpretation of the plant data as well as the availability of clear and unambiguous information and it is only when these features have been resolved is it possible examine the importance of differences in physical modelling. On this basis all the codes used demonstrated an ability to model gas-cooled systems under natural circulation and low pressure low Reynolds numbers' conditions.

Detection and Diagnostics of Loss of Coolant Accidents Using Support Vector Machines

It is very difficult for operators to predict the progression of a loss of coolant accident (LOCA) because nuclear plant operators are provided with only partial information during the accident or they may have insufficient time to analyze the data despite being provided with considerable information. Therefore, its break location should be identified and the break size should be predicted accurately in order to provide the operators and technical support personnel with important and valuable information needed to successfully manage the accident. In this paper, support vector machines (SVMs) are used to identify the break location of a LOCA and predict the break size using the support vector classification (SVC) and support vector regression (SVR), which are well-known application areas of SVMs. The SVR models to predict the break size were optimized using a genetic algorithm. The inputs to the SVMs are the time-integrated values obtained by integrating the measurement signals in a short time interval after a reactor scram. The results showed that the proposed algorithm identified the break locations of LOCAs without fault and predicted the break size accurately.

The multiple pressure tube rupture (MPTR) issue in RBMK safety technology

The RBMK core is constituted by more than one-thousand pressurized channels housed into stacked graphite blocks and connected at the bottom and at the top by small diameter ( D) and long length ( L) pipes (less than 0.01 and more than 10 m, respectively) that end up into headers and drum separators. Control valves are installed in the bottom lines. Due to the large L/ D value and to the presence of valves and other geometric discontinuities along the lines connecting with the pressure channels, the Fuel Channel Blockage (FCB) event is possible and already occurred in two documented NPP events. Pressure tube rupture occurred in a third NPP event not originated by FCB. Previous investigations, have shown the relevance of these events for the safety technology, and the availability of proper computational technique for the analysis, see the first and the third companion paper in this journal issue, respectively. The occurrence of the FCB event remains undetected for a few tens of seconds because of the lack of full monitoring for the individual channels, fourth companion paper in this journal issue. Therefore, fission power continues to be produced in the absence of cooling. This brings in subsequent times to fuel rod overheating, pressure tube failure, damage of the neighbouring graphite brick and ejection of damaged fuel. Following the pressure tube rupture, reactor cavity pressurization, radioactivity release into the same area and change of fluid properties occur that allow the detection of the event and cause the reactor scram at a time of a few tens of seconds depending upon the channel working conditions and the severity of the blockage. Notwithstanding the [delayed] scram and the full capability of the reactor designed safety features to keep cooled the core, the multiple pressure tube rupture (MPTR) issue is raised. The question to be answered is whether the ‘explosion’ of the blocked pressure tube damages not only the neighbour graphite bricks but propagates to other channels causing the potential for several channel failure. In order to address the MPTR issue fuel channel thermal-hydraulics and three-dimensional (3D) neutron kinetics analyses have been performed, as well structural mechanics calculations for the graphite bricks and rings (graphite rings surround the pressure tube to accommodate for thermal and radiation induced expansions). The bases for the analysis and the results of the study are presented. The conclusion, not reported within a licensing based format, is that the MPTR consequences are not expected to be relevant for the safety of the RBMK installations. This is supported by the analysis of experiments performed at the TKR facility available at the EREC research Centre near Moscow.

ICONE15-10489 An Enhanced Detect-and-Suppress Stability Protection Method for Boiling Water Reactors

Systems based on the Detect-and-Suppress (D&S) Long Term Stability Solution Option III have been installed in many BWR plants. Such systems issue scram signals upon detecting oscillatory neutron flux signals from Local Power Range Monitors (LPRM) grouped into several Oscillation Power Range Monitor (OPRM) cells. The licensing basis of the existing systems does not address situations leading to highly unstable conditions with multiple interacting instability modes. These conditions are anticipated with flow windows extended beyond the Maximum Extended Load Line Limit Analysis (MELLLA) domain, which are typically associated with power uprate. The Enhanced Option III (EO-III) solution presented in this paper is an evolutionary step relying on the existing methodology and hardware, and introducing measures for addressing the reduced stability associated with extended flow window conditions in general, and the higher probability of single channel hydraulic instability excitation in particular. The new elements introduced as enhancements to the existing Option III solution include the introduction of a calculated exclusion region on the power/flow map, protected by reactor scram, to preclude single channel instabilities. The calculation procedures for supporting the stability protection system are modified consistent with the introduction of the channel instability exclusion region. In particular, the limiting critical power response to power oscillations which is calculated using 3-D system codes, will always be guaranteed to be regular and bounded as a function of the oscillation magnitude when the flow in all the reactor channels oscillate coherently. In addition to the physical basis behind the improvements in the new EO-III, aspects of the engineering application and licensing will be also discussed.

Development of control technology for HTTR hydrogen production system with mock-up test facility: System controllability test for loss of chemical reaction

The Japan Atomic Energy Agency has been planning the demonstration test of hydrogen production with the High Temperature Engineering Test Reactor (HTTR). In a HTTR hydrogen production system (HTTR-H2), it is required to control a primary helium temperature within an allowable value at a reactor inlet to prevent a reactor scram. A cooling system for a secondary helium with a steam generator (SG) and a radiator is installed at the downstream of a chemical rector in a secondary helium loop in order to mitigate the thermal disturbance caused by the hydrogen production system. Prior to HTTR-H2, the simulation test with a mock-up test facility has been carried out to establish the controllability on the helium temperature using the cooling system against the loss of chemical reaction. It was confirmed that the fluctuations of the helium temperature at chemical reactor outlet, more than 200 K, at the loss of chemical reaction could be successfully mitigated within the target of ±10 K at SG outlet. A dynamic simulation code of the cooling system for HTTR-H2 was verified with the obtained test data.

RETRAN-3D analysis of the base case and the four extreme cases of the OECD/NRC Peach Bottom 2 Turbine Trip benchmark

This paper presents the results of RETRAN-3D calculations of the base case and the four extreme cases of phase 3 of the Peach Bottom 2 OECD/NRC Turbine Trip benchmark for coupled thermal-hydraulic and neutronic codes. The PSI-RETRAN-3D model gives good agreement with the measured data of the base case. In addition to the base case, the analysis of the extreme cases provides a further understanding of the reactor behaviour, which is the result of the dynamic coupling of the whole system, i.e., the interaction between the steam line and vessel flows, the pressure, the Doppler, void and control reactivity and power. For the extreme cases without scram the bank of safety relief valves is able to mitigate the effects of the turbine trip for short times. The 3-D nature of the core power distribution has been investigated by analysing the power density of the different thermal-hydraulic channels. In all cases prior to the reactor scram the course of the power is similar in all the channels with differences of the order of a few percent showing that, by and large, the core acts in a coherent manner. At the time of maximum power, the axial power distribution in the different channels is increased at the core centre with respect to the distribution at time zero, by an amount, which is different for the different channels.