Modelling of the spatial distribution of the induced activities in the RBMK-1500 reactor graphite blocks and rings/sleeves

There is one nuclear power plant (NPP) in Lithuania – the Ignalina NPP – which is under decommissioning now. The Ignalina NPP has two units with RBMK-1500 reactors, which are the most powerful and the most advanced versions of RBMK-type reactor design. Unit 1 of the Ignalina NPP was shut down at the end of 2004 and Unit 2 was shut down at the end of 2009. RBMK is a water-cooled graphite-moderated channel-type power reactor and the decommissioning of these reactors faces specific challenges for proper characterisation and disposal of irradiated reactor graphite.Apart from radiological inventory, the spatial distribution of radionuclides in the reactor graphite is also very important because it could indicate the possibilities for decontamination/treatment of the irradiated graphite. This is important for consideration of the near surface disposal option for irradiated graphite, as without treatment it usually does not meet the waste acceptance criteria.Based on that, the work presented in this paper is focused on the modelling of the induced activity spatial distribution in the Ignalina NPP RBMK-1500 reactor graphite components: blocks and rings/sleeves. The modelling was performed with MCNP and SCALE computer codes and consisted of two mains stages: modelling of the neutron flux in the reactor graphite components, and then modelling of the neutron activation in them using the already modelled neutron flux. In such a way, the spatial induced activity distribution in the analysed reactor components was obtained. Modelling results show that the thermal neutron flux is more intensive in the outer radial regions of the graphite components and this, in general, results in higher induced activities there.     

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.     

Analysis of potential waterhammer at the Ignalina NPP using thermal–hydraulic and structural analysis codes

The RBMK (Russian acronym for ‘channeled large power reactor’)-1500 reactors at the Ignalina nuclear power plant (NPP) have a series of check valves in the main circulation circuit (MCC) that serve the coolant distribution in the fuel channels. In the case of a hypothetical guillotine break of pipelines upstream of the group distribution headers (GDH), the check valves and adjusted piping integrity is a key issue for the reactor safety during the rapid closure of check valve. An analysis of the waterhammer effect (i.e. the pressure pulse generated by the valves slamming closed) is needed. The thermal–hydraulic and structural analysis of waterhammer effects following the guillotine break of pipelines at the Ignalina NPP with RBMK-1500 reactors was conducted by employing the RELAP5 and PipePlus codes. Results of the analysis demonstrated that the maximum values of the pressure pulses generated by the check valve closure following the hypothetical accidents remain far below the value of pressure of the hydraulic tests, which are performed at the NPP and the risk of failure of the check valves or associated pipelines is low. Sensitivity analysis of pressure pulse dependencies on calculation time step and check valve closure time was performed. Results of RELAP5 calculations are benchmarked against waterhammer transient data obtained by employing structural mechanics code BOS fluids.     

Ignalina NPP RBMK-1500 gas gap evaluation

This paper presents Ignalina NPP Unit 1 RBMK-1500 reactor core lifetime analysis. The closure of the gas gap between the pressure tubes and the graphite bricks is one of the criteria for the evaluation of the reactor core lifetime. The rate of closure of the approximately 1.5 mm gaps between the pressure tubes and the graphite is largely a function of accumulated fast neutron dose and graphite operating temperatures. The main task of this paper is development of strategy and methodology for gas gap closure evaluation.     

Analysis of BDBA in RBMK-1500 reactor with long-term loss of heat removal from the core

The Ignalina nuclear power plant (NPP) is a twin-unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. The accident management guidelines for beyond design basis accidents (BDBAs) are in a stage of preparation at Ignalina NPP. The most challenging event from BDBAs is the unavailability of water sources for heat removal from fuel channels (FCs). Due to specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: depressurisation of reactor cooling system (RCS) (if pressure in cooling circuit is high) and supply of water into cooling system from low pressure water sources, removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using cooling circuit of control and protection system channels, etc. The possibility to remove the heat using cooling circuit of control and protection system channels looks very attractive, because the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. The heat from fuel channels, where heat is generated, through graphite columns is transferred in radial direction to cooled channels with control rods. Therefore, the heat removal from RBMK-1500 reactor core using control rods cooling circuit can be used as non-regular mean for reactor cool-down in case of BDBAs with loss of long-term heat removal from the core.     

Development of accident management measures for RBMK-1500 in the case of loss of long-term core cooling

Results of the Level 1 Probabilistic Safety Assessment of the Ignalina Nuclear Power Plant have shown that in the risk topography transients are dominating. Analysis has shown that failure of the long-term core cooling is the main contributor to the core damage frequency. However, the reactor core damage in the long-term indicates the potential opportunities for the accident management. The main goal of accident management is to avoid a multiple fuel channel rupture because considering the design of RBMK reactors the consequences of rupture of more than 11–16 FC at full pressure would be close to the consequences of Chernobyl accident. This paper presents a detailed thermal–hydraulic analysis of the accidents with the loss of long-term core cooling, performed using the RELAP5 model of Ignalina NPP reactor cooling circuit and safety systems. Different ways of potential accident management are discussed. On the basis of this analysis the accident management strategy was developed.     

Specifics of RBMK core cooling in beyond design basis accidents

The most dangerous beyond design basis accidents for RBMK reactors, leading to the worst consequences, are related to the loss of long-term heat removal from the core. Due to a specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using control rods cooling circuit, depressurisation of reactor cooling system, supply of water into cooling system from low pressure water sources, etc. This paper presents the analysis of such heat removal by employing RELAP5, RELAP5-3D and RELAP/SCDAPSIM codes. The analysis was performed for Ignalina nuclear power plant with RBMK-1500 reactor. The analysis of result shows that the restoration of water supply into control rod channels enables to remove 10-30 MW of the generated heat from the reactor core. This amount of removed heat is comparable with reactor decay heat in long-term period and allows to slowdown the core heat-up process. However, the injection of water to reactor cooling system is considered as main strategy, which should be considered in RBMK-1500 accident management procedure.     

The individual channel monitoring (ICM) proposal to improve the safety performance of RBMK

A comprehensive study has been completed in relation to the accident analysis of RBMK with key results documented in three companion papers in the present journal volume. The study also aimed at the comparison between selected accident analysis topics in Light Water Reactors (LWR) and RBMK. The relevance of the Fuel Channel Blockage (FCB) event within the area of accident analysis was confirmed. Owing to the probability of occurrence, also connected with the large number of channels, estimated in the order of 10⁻²/reactor/year, the FCB is actually part of the Design Basis Accidents (DBA) for the RBMK.In case the FCB is part of the DBA, a noticeable difference occurs in results from safety evaluations between LWR and RBMK: in the latter case, a DBA causes a loss of integrity for the pressure barrier and, though to a limited extent, damaged fuel overpasses such a barrier. The FCB event also causes contamination in reactor cavity and in selected parts of the overall confinement, damage in various graphite blocks and, even excluding the MPTR risk owing to the findings from the second companion paper in this journal volume, mechanical loads on neighbouring fuel channels, graphite stacks and reactor tank that need extensive examination before reactor restart.Therefore, a proposal has been formulated for performing the feasibility analysis for the design of a system denominated ICM (Individual Channel Monitoring). The goal of the ICM is the early detection of the FCB event and the triggering of scram in such a way to prevent pressure tube damage and, definitely, over-passing of the pressure barrier by molten or damaged fuel during a DBA situation.The ICM is based upon the signals of pressure-drop (or flow-rate) and fluid temperature transducers installed in the bottom and the upper parts of the fuel channels, respectively. The performed study shows that steam superheating at fuel channel outlet occurs early after the blockage event and the related temperature signal can be used to cause scram. The availability of sophisticate computational tools including the detailed neutron kinetic model for each core channel, made possible the preliminary conclusion of the study.     

Characteristics of corium debris bed generated in large-scale fuel-coolant interaction experiments

The formation of corium debris as the result of fuel-coolant interaction (energetic or not) has been studied experimentally in the FARO and KROTOS facilities operated at JRC-Ispra between 1991 and 1999. Experiments were performed with 3–177 kg of UO₂–ZrO₂ and UO₂–ZrO₂–Zr melts, quenched in water at depth between 1 and 2 m, and pressure between 0.1 and 5.0 MPa. The effect of various parameters such as melt composition, system pressure, water depth and subcooling on the quenching processes, debris characteristics and thermal load on bottom head were investigated, thus, giving a large palette of data for realistic reactor situations.Available data related to debris coolability aspects in particular are:

• Geometrical configuration of the collected debris.

• Partition between loose and agglomerated (“cake”) debris.

• Particle size distribution with and without energetic interaction.

These data are synthesised in the present contribution.     

Qualification of the LF-eddy current technique for the inspection of stainless steel cladding and applications on the reactor pressure vessel

As part of the re-inspection of the reactor pressure vessel of the nuclear power plant, the low-frequency-eddy current technique was implemented during the 1995 outage. Since then, this inspection technique and the testing equipment have seen steady further development. Therefore, optimization of the entire testing system, including qualification based on the 1995 results, was conducted. The eddy current testing system was designed as a ten-channel test system with sensors having separate transmitter and receiver coils. The first qualification of the testing technique and sensors was performed using a single-channel system; a second qualification was then carried out using the new testing electronics. The sensor design allows for a simultaneous detection of surface and subsurface flaws. This assumes that testing is performed simultaneously using four frequencies. Data analysis and evaluation are performed using a digital multi-frequency regression analysis technique The detection limits determined using this technique led to the definition of the following recording limits for testing in which the required signal-to-noise ratio of 6 dB was reliably observed.

• Detection of surface connected longitudinal and transverse flaws:

• notch, 3 mm deep and 10 mm long, for weave bead cladding;

• notch, 2 mm deep and 20 mm long, for strip weld cladding.

• Detection of embedded planar longitudinal and transverse flaws:

• ligament of 7 mm for 8 mm clad thickness and 3 mm;

• ligament for 4 mm clad thickness, notch starting at the carbon steel base material with a length of 20 mm.

• Detection of embedded volumetric longitudinal and transverse flaws:

• 3 mm diameter side-drilled hole (SDH) for 8 mm clad thickness; ligament, 4 mm. For 4 mm clad thickness: diameter, 2 mm SDH; ligament, 2 mm. All SDHs are 55 mm deep.

The 2006 CHF look-up table

CHF look-up tables are used widely for the prediction of the critical heat flux (CHF). The CHF look-up table is basically a normalized data bank for a vertical 8 mm water-cooled tube. The 2006 CHF look-up table is based on a database containing more than 30,000 data points and provides CHF values at 24 pressures, 20 mass fluxes, and 23 qualities, covering the full range of conditions of practical interest. In addition, the 2006 CHF look-up table addresses several concerns with respect to previous CHF look-up tables raised in the literature. The major improvements of the 2006 CHF look-up table are:

• An enhanced quality of the database (improved screening procedures, removal of clearly identified outliers and duplicate data).

• An increased number of data in the database (an addition of 33 recent data sets).

• A significantly improved prediction of CHF in the subcooled region and the limiting quality region.

• An increased number of pressure and mass flux intervals (thus increasing the CHF entries by 20% compared to the 1995 CHF look-up table).

• An improved smoothness of the look-up table (the smoothness was quantified by a smoothness index).

A discussion of the impact of these changes on the prediction accuracy and table smoothness is presented. The 2006 CHF look-up table is characterized by a significant improvement in accuracy and smoothness.     

Validation of coupled neutronic/thermal-hydraulic code RELAP5-3D for RBMK-1500 reactor analysis application

This paper summarizes RELAP5-3D code validation activities carried out at the Lithuanian Energy Institute, which was performed through the modeling of RBMK-1500 specific transients taking place at Ignalina NPP. A best estimate RELAP5-3D model of the INPP RBMK-1500 reactor has been developed and validated against real plant data, as well as with the calculation results obtained using the Russian STEPAN/KOBRA code. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters, as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data, which demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors. Future activities are discussed.     

High performance light water reactor

The objective of the high performance light water reactor (HPLWR) project is to assess the merit and economic feasibility of a high efficiency LWR operating at thermodynamically supercritical regime. An efficiency of approximately 44% is expected. To accomplish this objective, a highly qualified team of European research institutes and industrial partners together with the University of Tokyo is assessing the major issues pertaining to a new reactor concept, under the co-sponsorship of the European Commission. The assessment has emphasized the recent advancement achieved in this area by Japan. Additionally, it accounts for advanced European reactor design requirements, recent improvements, practical design aspects, availability of plant components and the availability of high temperature materials. The final objective of this project is to reach a conclusion on the potential of the HPLWR to help sustain the nuclear option, by supplying competitively priced electricity, as well as to continue the nuclear competence in LWR technology. The following is a brief summary of the main project achievements:

• A state-of-the-art review of supercritical water-cooled reactors has been performed for the HPLWR project.

• Extensive studies have been performed in the last 10 years by the University of Tokyo. Therefore, a ‘reference design’, developed by the University of Tokyo, was selected in order to assess the available technological tools (i.e. computer codes, analyses, advanced materials, water chemistry, etc.). Design data and results of the analysis were supplied by the University of Tokyo. A benchmark problem, based on the ‘reference design’ was defined for neutronics calculations and several partners of the HPLWR project carried out independent analyses. The results of these analyses, which in addition help to ‘calibrate’ the codes, have guided the assessment of the core and the design of an improved HPLWR fuel assembly.

• Preliminary selection was made for the HPLWR scale, boundary conditions, core and fuel assembly design, reactor pressure vessel, containment, turbine and balance of plant.

• A review of potentially applicable materials for the HPLWR was completed and a preliminary selection of potential in-vessel and ex-vessel candidate materials was made.

• A thorough review of heat transfer at supercritical pressures was completed together with a thermal-hydraulics analysis of potential HPLWR sub-channels. This analytical tool supports the core and fuel assembly design.

• The RELAP5 and the CATHARE 2 codes are being upgraded to supercritical pressures. Thus they can be used to support the HPLWR core design and to perform plant safety analyses.

• Assessment of the HPLWR design constraints, based on current LWR technology was documented. This document stresses the various criteria that must be satisfied in the design (e.g. material, temperature, power, safety criteria, etc.) based on experience gained in the design of PWR.

• Preliminary economic assessment concluded that the HPLWR has the potential to be economically competitive. However, an accurate assessment can only be done after the HPLWR design has been fixed. A more accurate economic assessment may be performed after the conclusion of this project.

Modular HTR confinement/containment and the protection against aircraft crash

Does an HTR need a containment – pressure resistant – or is it possible – licensable – to have only a so-called confinement.The answer depends on both the results of the safety analysis of the accidents considered in the design and the acceptance by the licensing authorities and the public of a safety approach only based on severe core damage exclusion.The safety approach to be developed for modular HTRs must describe the application of the defence in depth principle for such reactors. Whatever the requirements on the last confinement barrier could be, a convincing demonstration of the exclusion of any severe core damage is needed, relying on exhaustive and bounding considerations of severe core damage initiators and the use of non-questionable arguments.The paper presents the containment issues for HTRs based on German experience background and considerations for modern modular HTR safety approach including beyond design situations.

• For the German HTRs (designed in the 80s), it could be shown in the licensing procedures in Germany that there was no need for a pressure retaining and gas tight containment to enclose radioactive nuclides released from the nuclear heat source. Instead, the confinement envelope acted in conjunction with other barriers to minimize the release of radioactive nuclides and the radiological impact on the environment.

• The confinement envelope consisted of the reactor building, a sub-atmospheric pressure system, a building pressure relief system, an HVAC systems isolation and a filtration system.

• During a major depressurization accident, unfiltered releases were discharged to the environment. The analyses results show that the environmental impact was far below the dose limits according to the German Radiological Protection Ordinance, even when the effect of filters was not taken into account.

• The demonstration strongly relied on the assumptions made for the source term definition, e.g. the fuel particles failure rates (under irradiation and during accidental conditions), the diffusion data, the dust data and the deposition/lift-off mechanisms.

• For modern modular HTRs, the last confinement barrier performances will have to be determined in accordance with the set of accidents to be considered in the design including internal and external hazards and the limits targeted for the public and the environment protection.

Further more the paper presents an analysis of effects of a deliberate crash of a large commercial airliner on a former German HTR design.     

SEMER: a simple code for the economic evaluation of nuclear and fossil energy-based power production systems

The code system, SEMER, was recently developed to evaluate the economic impact of various nuclear reactors and associated innovations. Models for nearly all fossil energy-based systems were also included to provide a basis for cost comparisons.Essentially, SEMER includes three types of model libraries: the global model, for a rapid estimation of various nuclear and fossil energy-based systems, the detailed models, for the finer cost evaluation of individual components and circuits in a PWR type of reactor and the fuel cycle models, for PWRS, HTRs and FBRs, allowing the cost estimations related to all the steps in the nuclear fuel cycle, including reprocessing and disposal.This paper summarises our on-going investigations on new developments in, and on the validation of, the SEMER system.Details of the modelling principles, and the results of validation carried out in the context of an EDF/CEA Joint Protocol Agreement, are also presented.First results of this validation are highly encouraging:

• Relative errors for the total kWh or overnight and investment costs are less than 5% for large PWR systems operating in France or other countries.

• These errors are less than 3% for small-sized compact PWRs and they are of the order of 4–7% for HTRs (as compared to IAEA estimations).

• For fossil energy-based power plants, the relative error, even with slightly different cost breakdown between SEMER and that of existing installations, is from 5 to 20%.

• Similarly, errors on the nuclear fuel cycle costs are about 1–4%, compared to published reference values.

Vessel and Channel Fast Reactors Cooled by Boiling Water or Water with Supercritical Parameters

The possibility is examined of developing a vessel-type fast reactor cooled by water with supercritical parameters (BR-VSP) and a channel-type fast plutonium–water reactor cooled by boiling water, using the RBMK scheme, or water with supercritical parameters for burning weapons or power-production plutonium (RBMK-Pu).A reactor with construction similar to that of an RBMK reactor but without the graphite moderator and zirconium can be used for burning plutonium. Removing the graphite masonry while retaining a 25-cm spacing in the square lattice gives a large free volume in the neutron field for holding, for example, cobalt. A neutron-physical substantiation is given for the serviceability of such a reactor with the maximum burnup of removed fuel 10% h.a. and the possibility of obtaining a negative reactivity effect in the case of water loss.     

A multi-tank storage facility to effect power control in the PBMR power cycle

This article presents the concept of a storage facility used to effect power control in South Africa's PBMR power cycle. The concept features a multiple number of storage vessels whose purpose is to contain the working medium, helium, as it is withdrawn from the PBMR's closed loop power cycle, at low energy demand. This helium is appropriately replenished to the power cycle as the energy demand increases. Helium mass transfer between the power cycle and the storage facility, henceforth known as the inventory control system (ICS), is carried out by way of the pressure differential that exists between these two systems. In presenting the ICS concept, emphasis is placed on storage effectiveness; hence the discussion in this paper is centred on those features which accentuate storage effectiveness, namely:

• Storage vessel multiplicity;

• Unique initial pressures for each vessel arranged in a cascaded manner; and

• A heat sink placed in each vessel to provide thermal inertia.

Having presented the concept, the objective is to qualitatively justify the presence of each of the above-mentioned features using thermodynamics as a basis.     

Relationship between nuclear graphite moderator brick bore profile measurement and irradiation-induced dimensional change

A stress analysis for a hypothetical nuclear graphite moderator brick is presented, considering dimensional and other property changes due to fast neutron irradiation, to illustrate the relationship between the change in moderator brick bore profile and dimensional change of the material. The results give the stresses and deformations of the brick during operation and at shutdown, with the effect of irradiation creep on the deformation of the brick also considered. The analyses provide information useful to reactor designers and operators for planning graphite monitoring campaigns.     

CEA and AREVA R&D on HTR fuel fabrication and presentation of the CAPRI experimental manufacturing line

In the framework of the French V/HTR fuel development and qualification program, the Commissariat à l’Energie Atomique (CEA) and AREVA are conducting R&D projects covering the mastering of UO₂ coated particle and fuel compact fabrication technology. To fulfill this task, a review of past knowledge, of existing technologies and a preliminary laboratory-scale work program have been conducted with the aim of retrieving the know-how on HTR coated particle and compact manufacture:

• The different stages of UO₂ kernel fabrication GSP process have been reviewed, reproduced and improved.

• The experimental conditions for the chemical vapor deposition of coatings have been defined on dummy kernels and development of innovative characterization methods has been carried out.

• Former CERCA compacting process has been reviewed and updated.

In parallel, an experimental manufacturing line for coated particles, named GAIA, and a compacting line based on former CERCA compacting experience have been designed, constructed and are in operation since early 2005 at CEA Cadarache and CERCA Romans, respectively. These two facilities constitute the CAPRI line (CEA and AREVA PRoduction Integrated line).The major objectives of the CAPRI line are:

• to recover and validate past knowledge,

• to produce representative HTR TRISO fuel meeting industrial standards,

• to permit the optimization of reference fabrication processes for kernels and coatings defined previously at a laboratory-scale and the investigation of alternative and innovative fuel design (UCO kernel, ZrC coating),

• to test alternative compact process options and

• to fabricate and characterize fuel required for irradiation and qualification purpose.

This paper presents the status of progress of R&D conducted on HTR fuel particles and compact manufacture by early 2005 and the potential of the laboratory-scale HTR fuel CAPRI line.     

Critical heat flux prediction in rod bundles under upward low mass flux densities

The analysis of experimental data and results of calculations for heat transfer crisis in heated channels under low upward coolant mass flux densities is presented. This analysis allows the determination of the basic features of the boiling crisis phenomenon. It is shown that the methods currently used for critical heat flux (CHF) prediction have insufficient accuracy in the given range of parameters. A new relationship for the CHF calculation is presented. It should be used for the water–water energy reactor (WWER) and uran–graphite channel reactor—Chernobyl-type (RBMK) rod bundles, and is verified by the test data. The comparison of results obtained by a new CHF correlation and the relationship used in RELAP5/MOD3.1 Code is presented. It is shown that the latter overpredicts the CHF values at atmospheric pressure and for x_cr>0.4 and does not provide conservative estimations for the RBMK fuel bundles.