Single rod experiments on transient void behavior during low-pressure reactivity-initiated accidents in light water reactors

A series of experiments for fast transient void behaviors during cold shutdown reactivity-initiated accidents in light water reactors has been performed with a single simulated fuel rod at the atmospheric pressure. A fast-response impedance technique was applied to the void fraction measurement. It was confirmed that, based on the comparison with X-ray technique, the impedance technique was capable of measuring void fraction within an allowable range of accuracy for fast transient two-phase flows. Experimental data was accumulated concerning the onset of net vapor generation, local void fraction, water temperature and pressure. The observed tendency on the onset of net vapor generation was qualitatively agreed with the Saha and Zuber model. However, the model overestimated local water subcooling at the onset of net vapor generation. Inlet water subcooling largely influenced the variation of local void fraction. It was observed that larger inlet water subcooling resulted in repeated growth and collapse of voids to suppress a continuous increase in void fraction.     

Review of thermal-hydraulic researches in severe accidents in light water reactors

The Tohoku Region Pacific Coast Earthquake and subsequent severe accident (SA) in Fukushima Daiichi Nuclear Power Station caused unprecedented disaster in Japan. Before this accident, considerable researches on SAs had been carried out in Japan. However, unfortunately, such researches could not prevent the accident due to the unexpected huge Tsunami. However, the researches on SAs become more and more important in order to make clear the causes of the accident in Fukushima and improve the safety of nuclear power plants in Japan. In view of this, review on researches on thermal hydraulics in SAs in light water reactors was carried out. Important thermal-hydraulic phenomena in SAs were identified. Research activities on each phenomenon were surveyed mainly based on the articles published in Journal of Nuclear Science and Technology of Atomic Energy Society of Japan.     

Optimisation of the operating conditions of pressurised water reactors limited by the DNB phenomenon

The replacement of steam generators in a pressurised water reactor (PWR) requires an adaptation of the normal operating conditions imposed by modified characteristics of the replaced steam generators (RSGs). To cover the cost of steam generator replacement, plant uprate is often considered. This is made possible by a significant improvement of the heat exchange characteristics of the SGs, and also by modification of the position of the turbine control valve.The present paper discusses the methodology that has been developed by Tractebel to solve this problem for most of the up-to-date PWRs that are limited by the departure from nucleate boiling (DNB) phenomenon (boiling crisis). A code has been developed, validated and applied to two units in Belgium, respectively Doel 3 and Tihange 1, for which it has been decided to replace the steam generators. This code is founded on the basic principle that for most DNB-limiting class-2 accidents, the maximum reactor power and reactor inlet temperature may not violate either the overpower limit (currently 118% of nominal power), or the DNB-ratio criterion in the course of this accident. The code allows the designer to assess the influence on the operating conditions of the various key safety parameters, as well as of the basic assumptions that are necessary to conduct the thermalhydraulic design work. Examples of such investigations are given in this paper. Special emphasis is put on the thermal-hydraulic (TH) design procedure, the plant principal TH key parameters, as well as the features of the new SGs. The development of the DNBOPT code was also motivated by the wish to perform audit and sensitivity calculations, as well as to confirm calculations performed by the SG supplier.     

Construction management of Indian pressurized heavy water reactors

Pandit Jawaharlal Nehru and Dr. Homi J. Bhabha, the visionary architects of Science and Technology of modern India foresaw the imperative need to establish a firm base for indigenous research and development in the field of nuclear electricity generation. The initial phase has primarily focused on the technology development in a systematic and structured manner, which has resulted in establishment of strong engineering, manufacturing and construction base.The nuclear power program started with the setting up of two units of boiling light water type reactors in 1969 for speedy establishment of nuclear technology, safety culture, and development of operation and maintenance manpower. The main aim at that stage was to demonstrate (to ourselves, and indeed to the rest of the world) that India, inspite of being a developing country, with limited industrial infrastructure and low capacity power grids, could successfully assimilate the high technology involved in the safe and economical operation of nuclear power reactors. The selection of a BWR was in contrast to the pressurized heavy water reactors (PHWR), which was identified as the flagship for the first stage of India's nuclear power program. The long-term program in three stages utilizes large reserves of thorium in the monazite sands of Kerala beaches in the third stage with first stage comprising of series of PHWR type plants with a base of 10,000 MW. India has at present 14 reactors in operation 12 of these being of PHWR type.The performance of operating units of 2720 MW has improved significantly with an overall capacity factor of about 90% in recent times.The construction work on eight reactor units with installed capacity of 3960 MW (two PHWRs of 540 MW each, four PHWRs of 220 MW each and two VVERs of 1000 MW each) is proceeding on a rapid pace with project schedules of less than 5 years from first pour of concrete. This is being achieved through advanced construction technology and management. Present efforts are focused on further reduction of gestation period. This is in contrast to construction period of 7–14 years in the earlier projects with labour intensive construction methods, learning period and indigenisation. The schedule and cost are interrelated and ultimately determine the viability and competitive edge of a project. With rich experience of over 30 years of operation and construction management it is well established that setting up of nuclear power projects in India in 4–5 years is quite feasible because of tremendous developments in construction technology; mechanization, parallel civil works and equipment erection, computerized project monitoring and accounting systems. By considering the best achieved times for the critical path activities of previous and ongoing projects, even a 4-year schedule is achievable. For nuclear power to be competitive it is essential that the gestation period is reduced and the capacity utilization enhanced. Both of these are the goals of the Indian nuclear power program. Presently the overnight cost per kW installed capacity is in the range of US$ 1100–1300 with levellised tariff of 5 c/kWh.     

Unanticipated accidents in sodium-cooled fast reactors

The definitions and requirements of normative documents for unanticipated accidents at nuclear power plants with fast reactors are analyzed. Definitions are constructed between one another and with a collection of scenarios which can lead to unanticipated accidents, likewise determined by normative documents independently of the probability of these accidents actually happening. It is concluded that the normative approaches to fast-reactor safety must be refined with respect to strengthening the probabilistic criteria as a tool limiting the list of required unanticipated accidents for validating reactor safety. Special attention is devoted to the need to strengthen the motivation of designers to make the maximum possible use of passively triggered safety systems.     

Investigation of material relocation phenomena during severe accidents in fast breeder reactors

The SIMBATH out-of-pile experiments simulate severe accidents in fast breeder reactors. In the tests the nuclear energy released is substituted by the exothermal energy of a thermite reaction. Single pin and small bundle experiments as well as freezing tests are performed. Material ejected from the fuel rod simulators in an early phase is finely dispersed. A portion penetrates the upper breeding zone without freezing. The bulk of molten material ejected afterwards leads to blockages in the colder zones of the bundle. Under these conditions bottled-up situations may occur in the SIMBATH experiments. Residual sodium may become entrapped. The current version of the computer code CALIPSO developed to interpret these experiments is verified by calculation of two single pin experiments. The computations show that the relocation mechanisms in the SIMBATH experiments are mainly controlled by expansion of noncondensible gases originally existing inside the pins. The contribution from fuel vapour pressure or from sodium evaporation due to fuel-coolant-interaction is of less importance during the first 100 ms after fuel pin failure.     

Operating experience with pressurised components of German light-water reactors

Operating experience of pressurised components is reported on the basis of 19 light-water reactors operating in Germany. The design basis and materials have demonstrated their worth. Licenses are not limited in time, and the major regulatory effort is directed to continuous improvement in plant safety. Technical issues for long-term operation as evaluation of operating experience, plant monitoring, replacement of components are addressed. The basic safety concept as the design basis for the pressurised components is illustrated by some of its details. The main results from the analysis of operating experience are mentioned. Plant monitoring and inspections are important measures to ensure the integrity of the components and to maintain the safety level of the plant. The expansion of the monitoring system may allow a reduction in the scope of inspections.     

Intelligent system for transient data collection and fatigue monitoring of pressurised water reactors nuclear steam supply system

Electricité de France (EDF), the French national electricity company, is operating 54 standardised pressurised water reactors. This about 500 reactor-years experience in nuclear stations operation and maintenance area has allowed EDF to develop its own strategy for monitoring of age-related degradations of NPP systems and components relevant for plant safety and reliability. After more than fifteen years of experience in regulatory transient data collection and seven years of successful fatigue monitoring prototypes experimentation, EDF decided to design a new system called SYSFAC (acronym for SYstème de Surveillance en FAtigue de la Chaudière) devoted to transient logging and thermal fatigue monitoring of the reactor coolant pressure boundary. The system is fully automatic and directly connected to the on-site data acquisition network without any complementary instrumentation. A functional transient detection module and a mechanical transient detection module are in charge of the general transient data collection. A fatigue monitoring module is aimed towards a precise surveillance of five specific zones particularly sensible to thermal fatigue. After a first step of preliminary studies, the industrial phase of the SYSFAC project is currently going on, with hardware and software tests and implementation. The first SYSFAC system will be delivered to the pilot power plant by the beginning of 1996. The extension to all EDF’s nuclear 900 MW is planned after one more year of feedback experience.     

Acoustic emission technique for leak detection in an end shield of a pressurised heavy water reactor

This paper discusses the successful application of the Acoustic Emission Technique (AET) for detection and location of leak paths present on the inaccessible side of an end shield of a Pressurised Heavy Water Reactor (PHWR). The methodology was based on the fact that air and water leak AE signals have different characteristic features. Baseline data was generated from a sound end-shield of a PHWR for characterizing the background noise. A mock up end-shield system with saw cut leak paths was used to verify the validity of the methodology. It was found that air leak signals under pressurisation (as low as 3 psi) could be detected by frequency domain analysis. Signals due to air leaks from various locations of a defective end-shield were acquired and analysed. It was possible to detect and locate leak paths. Presence of detected leak paths were further confirmed by alternate test.     

Effect of processing on properties of thin walled calandria tubes for pressurised heavy water reactor

Thin walled calandria tubes for pressurised heavy water reactors are manufactured either by seam welding of Zircaloy-4 sheets or by seamless route. In the present study, the effect of processing on the critical properties such as texture, microstructure, hydriding behaviour and residual stress for both the routes as well as the mechanical anisotropy developed due to seam welding are investigated. The properties of the seam welded tube in the fusion and adjoining region are markedly different from the base material and from the seamless tube. Residual stress measurements indicate that heat affected zone (HAZ) of seam welded tubes have longitudinal tensile residual stress and the seamless tubes have uniform compressive stress along the circumference. The phase transition in the presence of residual stresses due to thermal gradient is found to modify the texture in the HAZ. The hydride orientation and mechanical anisotropy in these regions are found to be dependent on the texture of the material.     

Thermo-hydraulics during start-up in natural circulation boiling water reactors

The transient behavior of natural circulation for boiling two-phase flow was investigated by simulating normal and abnormal start-up conditions to research the feasibility of natural circulation BWRs such as the SBWR. It was found that the instabilities, which are out-of-phase geysering, in-phase natural circulation oscillation and out-of-phase density wave instability, may occur during the start-up when the vapor generation rate is insufficient. In this paper, the mechanism of in-phase natural circulation oscillation induced by hydrostatic head fluctuation in steam separators, which has never been understood well enough, is experimentally clarified. Next, the effect of system pressure on the occurrences of the geysering and the natural circulation oscillation are investigated. Finally, from the results, a recommendation is provided to establish the rational start-up procedure and reactor configuration for natural circulation BWRs.     

Plutonium recycle in heavy water-power reactors

The performance of a heavy-water reactor using U²³⁸ with an equilibrium concentration of U²³⁹ Pu²⁴⁰ Pu²⁴¹ nuclei and small additions of U²³⁵ and with natural and depleted uranium make-up is discussed. A portion of the spent fuel discharged from the reactor is cleaned up of fission fragments and recycled to the reactor. Another (smaller) portion is withdrawn from the cycle after plutonium has been extracted and is replaced in the core by natural uranium.Translated from Atomnaya Énergiya, Vol. 20, No. 1, pp. 26–29, January, 1966     

Fuel-element simulator for investigating thermal-hydraulic accidents in water-water reactors

Power Physics Institute, Obninsk. Translated from Atomnaya Énergiya, Vol. 73, No. 6, pp. 470-474, December, 1992.     

Analysis of hexcan failures occurring during the simulation of severe accidents in liquid metal cooled reactors

Under the SIMBATH programme the physical phenomena of transient material movement and relocation during severe LMFBR accidents are investigated out-of-pile. In most of the SIMBATH bundle experiments a failure of the wrapper was observed. From the safety point of view this has implications on the issue of propagation. By openings into the inter-subassembly gaps pressure relief and material release are possible. From the development of failure, based on measurements made during the simulation tests, and from post-experiment investigations three types of failure mode have been identified:

• - Melt-through of the wrapper wall by a jet of hot material from a failing pin. This happened very early during the test. Sodium boiling in the annular bypass prior to failure has not been detected.

• - Melt-through in the simulated fuel region by severe ablation due to local crust instability combined with intense heat input from the flowing melt.

• - Melt-through in the simulated breeding regions close to blockages. This failure mode was always observed together with sodium gross boiling in the annular channel, i.e. reduced cooling of the wrapper wall.

No mechanical failure was detected as a result of the stress concentration in the corners of the hexcan walls. The influence of the internal overpressure is restricted mainly to final break-through after severe ablation and drives the material motions after wrapper failure; it does not control wrapper wall failure in these experiments.     

Counter-current flow limitations during hot leg injection in pressurized water reactors

A one-dimensional model is presented to predict counter-current flow limitations during hot leg injection in pressurized water reactors. Different from previous models, it may also be applied in case of high Froude numbers of the liquid flow, such as to be expected in the case of emergency coolant injection through the hot leg. The model has been verified with an extensive experimental program performed in the WENKA test facility at the Forschungszentrum Karlsruhe. Typical flow regimes were investigated for a wide range of flow conditions, simulated with air and water at ambient pressure and temperature, in a simplified Pressurized Water Reactor (PWR) hot leg geometry. Depending on the water and air flow rates, flow phenomena such as a hydraulic jump and flow reversal were experimentally observed. The theoretical model shows that not only the nondimensional superficial velocities of liquid and gas, but also the Froude number of the liquid at the injection point and the Reynolds number of the gas play an important role for the prediction of flow reversal. In case of a high liquid inlet Froude number, a flow reversal could only be observed if the liquid flow became locally subcritical, i.e. if a hydraulic jump occurred in the channel. The flow reversal is predicted by the presented model with good accuracy.     

Features of major accidents in VPBÉR-600 reactors

Materials Safety Design Office. Translated from Atomnaya Énergiya, Vol. 74, No. 4, pp. 294-302, April, 1993.     

Impact of different moderator ratios with light and heavy water cooled reactors in equilibrium states

As an issue of sustainable development in the world, energy sustainability using nuclear energy may be possible using several different ways such as increasing breeding capability of the reactors and optimizing the fuel utilization using spent fuel after reprocessing as well as exploring additional nuclear resources from sea water. In this present study the characteristics of light and heavy water cooled reactors for different moderator ratios in equilibrium states have been investigated. The moderator to fuel ratio (MFR) is varied from 0.1 to 4.0. Four fuel cycle schemes are evaluated in order to investigate the effect of heavy metal (HM) recycling. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of SRAC 2000 code using nuclear data library from the JENDL 3.2. The results show a thermal spectrum peak appears for light water coolant and no thermal peak for heavy water coolant along the MFR (0.1 ? MFR ? 4.0). The plutonium quality can be reduced effectively by increasing the MFR and number of recycled HM. Considering the effect of increasing number of recycled HM; it is also effective to reduce the uranium utilization and to increase the conversion ratio. trans-Plutonium production such as americium (Am) and curium (Cm) productions are smaller for heavy water coolant than light water coolant. The light water coolant shows the feasibility of breeding when HM is recycled with reducing the MFR. Wider feasible area of breeding has been obtained when light water coolant is replaced by heavy water coolant.     

Core coolability of an ATR by heavy water moderator in situations beyond design basis accidents

In the case of a loss-of-coolant accident (LOCA) with coincident loss of emergency coolant injection (LOECI), core cooling is generally very severe. However, as the ATR plant has heavy water at about 60°C in the core, decay heat can be removed by the heavy water cooling system. Separate-effects tests relating to heavy water cooling were conducted with each setup. The important thermal hydraulics was radiation heat transfer, ballooning of a pressure tube, contact conductance between the pressure tube and a calandria tube and critical heat flux of the calandria tube. Constants and correlations obtained by the tests were incorporated into several codes to assess the core cooling. Long term core cooling capability with the heavy water cooling system was assessed. The core was cooled without melting under the postulated events due to inherent characteristics of the ATR.