Determining the three-dimensional energy output field in an RBMK-1500 fuel assembly

Ignalina Nuclear Power Plant. Translated from Atomnaya Énergiya, Vol. 70, No. 4, pp. 246–248, April, 1991.     

Approach to accident management in RBMK-1500

In order to ensure the safe operation of the nuclear power plants accident management programs are being developed around the world. These accident management programs cover the whole spectrum of accidents, including severe accidents. A lot of work is done to investigate the severe accident phenomena and implement severe accident management in NPPs with vessel-type reactors, while less attention is paid to channel-type reactors CANDU and RBMK.Ignalina NPP with RBMK-1500 reactor has implemented symptom based emergency operation procedures, which cover management of accidents until the core damage and do not extend to core damage region. In order to ensure coverage of the whole spectrum of accidents and meet the requirements of IAEA the severe accident management guidelines have to be developed.This paper presents the basic principles and approach to management of beyond design basis accidents at Ignalina NPP. In general, this approach could be applied to NPPs with RBMK-1000 reactors that are available in Russia, but the design differences should be taken into account.     

Calculating coolant inlet temperatures for VVÉR fuel-pin assemblies

Translated from Atomnaya Énergiya, Vol. 68, No. 3, pp. 206–207, March, 1990.     

Updating of risk-informed ISI programme for Ignalina RBMK-1500 nuclear power plant in Lithuania: Results and challenges

The paper presents a risk-informed in-service inspection (RI-ISI) pilot study project of 300 mm piping at Ignalina nuclear power plant (INPP) RBMK-1500 reactor, located in Lithuania. The RI-ISI study investigates optimal 300 mm piping ISI strategies with respect to risk and required resources. In total 1240 stainless steel welds were analyzed, assuming inter-granual stress corrosion cracking (IGSCC) to be the main damage mechanism. Pipe break frequency was estimated by probabilistic fracture mechanics methods and combined with safety barriers, provided by probabilistic safety assessment (PSA) study.After 3 years of operation, updating of RI-ISI was performed by taking into account new statistical data on pipe defects. Comparison with previous RI-ISI program was performed. The paper includes discussion on uncertainties in the study and robustness of RI-ISI programs.     

Improvement of the method used for hydraulic profiling of an RBMK-1500 core at the Ignalina nuclear power plant

An improvement of the method of hydraulic profiling of the RBMK-1500 core at the Ignalina nuclear power plant is examined. The hydraulic profiling method employs an approach where an adequate margin up to a crisis of heat transfer is provided in the channels of the reactor, the mass steam content at all channel exits is kept uniform, the number of adjustments of the stop-control valve at power is kept to a minimum, the flow rate is changed three times over the entire run of a fuel assembly, and the total flow rate through the reactor is maintained within the regulation limits over the entire run. The GIDRA software developed makes it possible to perform hydraulic profiling of RBMK-1500 using the method indicated. __________ Translated from Atomnaya énergiya, Vol. 101, No. 5, pp. 342–346, November, 2006.     

Determination of I activity in the RBMK-1500 main circulation circuit

The purpose of this work was to evaluate the content of difficult to measure isotope ¹²⁹I in the RBMK-1500 reactor fuel-to-clad gap and reactor main circulation circuit (MCC) coolant. To determine fission product (FP) release from the defective fuel, the methodology proposed by Lewis and Husain for the CANDU reactor primary coolant activity prediction was applied. The determined effective diffusion coefficient D′ = 1.2E−09 s⁻¹ of iodine in the RBMK-1500 fuel is higher than the one evaluated for the CANDU fuel 6.8E−10 s⁻¹. Results show that the method developed by Lewis and Husain can be applied for the RBMK-1500 fuel gap and reactor main circulation circuit coolant activity prediction.     

Evolution of the neutron sensor characteristics in the RBMK-1500 reactor neutron flux

Variation of characteristics of the RBMK-1500 reactor radial neutron flux sensors with the HfO₂ emitter during long-term maintenance was investigated. The influence of nuclear fuel enrichment and burnable erbium admixtures on the energy neutron spectrum, neutron absorption, and hafnium isotopic composition variation was considered. The dependences of corrective factors of the neutron sensor signal on the nuclear fuel burnup depth and the integral current accumulated by the sensor for different enrichment nuclear fuel are presented in the work. The experimental verification of the calculated dependence of the sensor corrective factor on the accumulated integral current was performed.     

Study of the nuclide inventory of operational radioactive waste for the RBMK-1500 reactor

The experimental and modeling results on the radionuclide concentrations in the Ignalina NPP operational waste are presented in the work. The scaling factors between easy-to-measure γ emitters ¹³⁷Cs, ⁶⁰Co and a number of difficult-to-measure radionuclides, the activity measurements of which are related to radiochemical procedures, α and β spectrometry, have been determined. The study shows that the scaling factor method can be applied for RBMK-1500 reactor waste characterization. The scaling factors were used in determination of the nuclide composition of operational radioactive waste and characterization of radioactive waste during the Ignalina NPP decommissioning.     

Safety analysis of beyond design basis accidents in RBMK-1500 reactors

At present the design basis accidents for RBMK-1500 are rather thoroughly investigated. The performed analyses helped to develop and implement a number of safety modifications. Further plant safety enhancement requires developing emergency procedures that would enable beyond design basis accidents management by preventing core damage or mitigating consequences of severe accidents.     

Revised calculation route for two group macroscopic cross-section library for RBMK-1500 type reactor

To increase the accuracy of predicted reactivity effects and coefficients for the unit equipped with a RBMK-1500 type reactor at Ignalina NPP, the calculation route used to generate the library of nuclear data constants applied in the neutronic/thermal hydraulic analysis has been updated with a modern version of the WIMS lattice code, WIMS8. The previously available two group library used with the QUABOX/CUBBOX-HYCA code, adapted to model the physical and nuclear processes in a RBMK-1500 reactor core, was created using the freely available WIMSD reactor physics cell code and its associated nuclear data library. In this article, the results that are obtained under the performance of the new two group cross-section library generated with WIMS8 for RBMK-1500 design core are presented. This discussion is mostly concentrated on the prediction of the key physics parameter for the RBMK type reactor, the void reactivity coefficient, as this parameter has been underestimated, especially at higher fuel irradiation.     

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.     

Integral analysis of large LOCA beyond design basis accident scenario in RBMK-1500

The integral analysis of severe accident scenario for RBMK-1500 was performed using combined approach with RELAP5, RELAP/SCDAPSIM, ASTEC and COCOSYS codes. The performed analysis covered response of the reactor core, the reactor cooling system and the confinement. There were performed several analyses: the first analysis assumed that operators take no action or their actions are not successful to provide the coolant injection to the reactor core; the other analyses were performed to investigate the accident management measures to restore the core cooling at different temperatures of the reactor core. The results of performed analyses showed that the operators have ∼5 h before the ruptures of fuel claddings occur and ∼8 h before the onset of exothermic steam-zirconium reaction. The coolant injection to the reactor core should be restored as soon as possible in order to prevent high hydrogen concentrations in the confinement and significant release of the fission products to the environment.     

Analysis of relative release rates of radionuclides from the RBMK-1500 reactor fuel elements

Specific activities (concentrations) of fission products (FP) and activation products in spent fuel elements of the RBMK-1500 reactor were calculated using SCALE 5 computer code. Different burnup (5.1–21.0 MWd/kg) fuel assemblies were experimentally investigated. Activities of radionuclides present in the coolant water of storage cases of defective fuel elements were experimentally measured and analyzed. Experimental results provide a basis for a quantitative analysis of radionuclide release from spent fuel of the RBMK-1500 reactor. Relative release rates of radionuclides from the fuel matrix were assessed based on a comparison of experimental results with theoretical calculations. On the basis of analysis results released fission and activation products can be divided into several groups according to their release rates from fuel; this can be generalized for radionuclides with similar chemical properties.