The reactor trip in emergency is done by simultaneous insertion of the control rods into the core by gravity following the drive mechanism de-energization, which is actuated by trip signals from the automatic control system. In the

case of failure to actuate the electromechanical protection system, the reactor shutdown is accomplished by the emergency boron injection system. Activation of the system is done by manually opening valves in the pipelines connecting the system to the reactor. Both shutdown systems ensures the reactor shutdown and its shutdown margin is sufficient enough to keep the cold clean reactor in a subcritical state.

2.3.2 Residual Heat Removal System

Normal decay heat removal when cooling down for maintenance and refueling, the steam generators with turbine bypass system are used and heat is rejected through the condensers. This can be achieved by natural circulation on the primary side but requires feed pumps and other equipments on the secondary system. If the secondary system is not available, active decay heat removal system with steam generators are used to remove decay heat and heat is rejected through the component cooling system.

Should there be no ac power available, decay heat is removed by natural convection system which only requires battery power to operate the initiation valves and passive decay heat removal system which is composed of heat pipe and heat exchangers. The heat is ultimately rejected to atmosphere by natural convection of air flow. Thus, there is theoretically infinite time of heat removal without operator intervention. One of the advantages of heat pipe passive decay heat removal system is that this system is continuously operating during normal plant operation to remove the heat loss from reactor vessel through the wet thermal insulation.

The application of supercritical pressure in both thermal and fast

pressurized water power reactors with thermodynamic efficiency up to 44% was considered in many designs in the 60’s. The main difficulties in the construction of such reactors were connected with the development of reactor materials required for work under high temperature conditions, for example, for the efficiency of 44% it was necessary to have a primary coolant temperature of about 550°C. The SCP LWR construction was postponed because of this problem and good perspectives for the development of LWRs with subcritical pressure.

The RSC Kurchatov Institute returned to the idea of the NPP

development with SCP at the beginning of the 80’s. After severe accidents at NPPs the accent was transferred from the increase of NPPs efficiency to enhancing the safety while keeping economic competitiveness. The RSC

Kurchatov Institute has carried out studies which show that in the case of pressure increase over the critical value in an integral LWR all safety requirements for the future reactors can be met with economic characteristics at least on the level of NPPs of traditional layout.

The following four main requirements formed the basis of the

development of the reactors at SCP carried out jointly by the RSC Kurchatov Institute and EDO Hydropress. They are:

— its safety level must meet safety requirements for the reactors of future generation,

— its economic characteristics must have advantage over other reactors,

— it must have potentially lower environmental impact,

— it should not depart significantly from existing LWR technology and take into

consideration the utilization experience of the water at SCP in heat power stations.

It must be noted, that nowadays water at SCPs is widely used in fossil plants. The operation of these power plants has shown a high reliability of the equipment under these conditions.

Fig. l shows water enthalpy and density at SCP 23.6 MPa (Pcr=22.1 MPa) versus temperature. The temperature at which the derivative of enthalpy versus temperature is maximum is denoted as Tm. For comparison physical properties of water at 15.7 MPa are given in Fig. l.

2800

tuO

Л

З 2300 J a

л 1800

r

1300

300 320 340 360 380 400

Temperature °С, T

Fig. 1. Water enthalpy and density vs. temperature.

There are some correlations, e. g. V. Protopopov, V. Silin [2,3] for the calculation of heat transfer of SCP water flow inside the tubes and correlations

for the conditions at which heat transfer deterioration was observed.

The large amount of heat transfer experimental data of SCP water flow

in large bundles was obtained in the RSC Kurchatov Institute.

The experimental ranges covered are:

Pressure: 23.5 and 29.4 MPa,

Mass velocity: 350 to 5000 kg/(m2/c),

Bulk water enthalpy: 1.0 to 3.0 MJ/kg,

Heat flux: 0.18 to 4.5 MW/m^.

Experimental heat transfer was satisfactorily described by correlations obtained at SCP water flow in tubes as a resulting. Note that in the experiments conducted in the RSC Kurchatov Institute there was no heat transfer deterioration in the experiments with multirod bundles within the same test parameter range at which heat transfer deterioration occurred in tubes. Available information on supercritical heat transfer and hydrodynamics allowed reliable estimation of thermohydraulics characteristics of the core, SG and primary loop.

The value of primary operation pressure (P0) was determined from the requirement to maintain the supercritical pressure during transients taking into account pressure support system features. The minimal margin relative to

critical pressure Pcr was adopted equal to 1.5 MPa which defined the PD value equal to 23.6 Mpa. A greater pressure margin would cause increase in vessel mass and consequently would worse the technical and economic characteristics. The coolant temperature was chosen on the basis of the following ideas:

— thermodynamic efficiency increases with the temperature but if it is above the Tm value it is more difficult to supply the NPP with necessary materials and the water properties of as a coolant become worse,

— to improve the fuel cycle features and safety level a decision was made to use

the sharp change of the coolant density in the vicinity of Тщ temperature to maintain the core criticality during fuel lifetime.

Taking into account these considerations the core inlet temperature was chosen below Tm and the core outlet temperature was chosen close to Tm that is approximately 380°C.

To maintain the core criticality between refuelings the coolant density increases smoothly during fuel lifetime. This is achieved by primary coolant temperature decrease in the SG which increases the feedwater flow at the given reactor thermal power. The growth of feedwater flow decreases the steam overheating in the SG and increases the heat transfer to the secondary side. As a result the primary coolant temperature decreases at SG outlet leading to the decrease of core inlet and core average coolant temperatures.

The chosen coolant parameters provide:

— growth of NPP efficiency up to 38%,

— several times decrease of SG heat transfer specific surface due to the increase

of temperature difference between the primary and secondary sides as compared to an NPP with subcritical pressure,

— decrease of coolant mass flow due to the high coolant heat capacity in the region of Tm,

— significant difference of coolant densities at core inlet and outlet,

— use of a fuel cycle with high conversion coefficient and core criticality maintenance by varying the neutron spectrum during fuel lifetime due to the sharp density increase in the region of Tm temperature,

— high heat transfer coefficients in the core and SG,

— high average coolant heat capacity.

Taking these factors into account it was decided to develop the design of reactor of maximum unit power with natural circulation with regard to the capabilities of the existing reactor technology. The parameters of steam fed to the turbine at the chosen coolant temperature were adopted on the basis of optimization calculations. In these calculations the NPP efficiency increase with pressure and steam overheating and simultaneous decrease of average specific

heat flux in the SG were taken into account The decrease of specific heat flux in the SG requires the increase of SG heat transfer surface at given thermal reactor power.

S. G. KALI AKIN, Yu. P. DZHUSOV,

R. V. SHUMSKY, Yu. STEIN Institute of Physics and Power Engineering,

Obninsk, Kuluga Region,

Russian Federation

Abstract

The results of the investigation of siectm-zirconinm reaction kinetics at the HPE simulator surface are presented in the paper. The dynamic characteristics of the hydrogen production resultedfrom the healed surface dry out are determined.

INTRODUCTION

The necessity of the knowledge of the laws of hydrogen generation during the interaction between the HPE shells made of the zirconium alloy and water coolant is highly increased when the problem of severe out-of-project accident was initiated, so their evolution is determined by the hidrogen discharge.

In [1] on the base of publications available the analisis was made of the hidrogen generation processes under severe accidents resulted from the circulation break off. From the estimations presented there regarding the amount of generated hydrogen during some first hours after the accident, it follows that the general source of the hydrogen is the reaction of zirconium with steam-water mixture. For example, the velocity of hydrogen generation may reach the value of 726 kg/h for the ordinery commercial APP USA.

Among the latter publications dealing with this problem the paper of the Japan authors 12] is remarcable. In this paper the results of the in-pile experiments are presented devoted to the study of hydrogen generation during the steam-zirconium reaction when the accidant situation was initiated by reactivity. The HPE was cooled by subsaturated water in the film boiling regime. The amount of the hydrogen generated was determined by the void fraction gauge at the channel exit and from the metallografic examinations also. In spite of the significant discrepancy between the results obtained by these two methods, the data allow to make conclusions about the strong temperature influence upon the hydrogen production.

Investigation.

1 — Heating Element (HPE Simulator); 2 — Transparent Body; 3 — Bars; 4 — Фланец; 5 — Thermocouples; 6 — Refrigerator; 7-Hydrogen Accumulator; 8 — Level Gauge SAPPHIRE-22DD Type; 9 — Replacing Tank; JO — Water Feed Tank.

The aims of the present paper are as follows: experimental study of the hydrogen generation process as a result of the HPE zirconium shell oxigation in steam-water atmosphere under conditions of assembly dryout due to the circulation failure; using of the data obtained when developing the corresponding codes.

Presented in the paper are the results of first experiments having, in general, methodological purposes. At the one-rod assembly model the peculiarities of the hydrodynamic processes of its drv-out was visualised and the reliable method of hydrogen production measurement was veryfied.

Integral reactors, developed in OKBM, being one of the varieties of PWR, are based on the common research and development work and on the experience in the creation, operation and development of such reactors.

But the novelty of the design decisions, connected with the integral lay-out of the reactor, the presence of a steam-gas pressurizer and guard vessel, the absence of circulation loops in the circuit and some others, demands special research work to be performed.

A lot of research work, connected with the experimental study of thermo-hydraulic processes in integral PWRs with a built-in steam-gas pressurizer have been performed in the existing experimental base.

The experimental investigations performed confirmed the main design decisions for equipment and systems and allowed substantion of the correctness of the chosen regime parameters, reliability and safety of the plant.

Together with the problems of the study of thermo-hydraulic processes, occuring in the plant in emergency conditions and of substantiation of the operability and efficiency of the provided safety systems, the most important problem for the experiments is to collect representative information for computer code verification.

The main investigations, which are being performed at present are the following:

— investigations of DNB in fuel assemblies and temperature state of fuel elements at partial and complete dry-out of the core;

— investigation of the conditions of steam condensation from steam-gas mixture in the heat exchangers-condensers of the emergency residual heat removal systems and in the built-in steam generators;

— investigations of steam-gas mixture distribution inside the pressurizer;

— investigations of water-gas and chemical conditions in the primary circuit, including gas transfer in the circuit;

— investigations at integral facilities, including a wide range of emergency conditions with primary circuit loss of tightness and heat removal disruption;

— investigations, verifying thermo-hydraulic and lifetime characteristics of steam generators.

Fig.4 shows the complex of facilities for thermo-physical investigations and safety of VPBER-600. *

To verify the results on the problem of corium confinement, it is necessary to perform additional investigations into thermo-physical, physico-chemical and thermo-mechanical processes, to improve calculation modes and computer programs. Besides, the conservativeness of assumptions made considerably compensates for the lack of information and gives every reason to obtain a positive solution of the problem of corium confinement in the reactor vessel or guard vessel.

Now testing facility for an integral PWR of 200 MW power is being made ready for putting into operation.

A. R. ANTARIKSAWAN, I. SUBKI National Atomic Energy Agency,

Tangerang, Indonesia

Abstract

Based on the result of the Feasibility Study undertaken since 1991, Indonesia may enter in the new nuclear era by introduction of several Nuclear Power Plants in our energy supply system. Requirements for the future NPP’s are developed in two step approach. First step is for the immediate future that is the next 50 years where the system will be dominated by A-LWR’s/A-PHWR’s and the second step is for the time period beyond 50 years in which new reactor systems may start to dominate. The integral reactor concept provides a revolutionary improvements in terms of conceptual and safety. However, it creates a new set of complexe machinery and operational problems of its own. The paper concerns with a brief description of nuclear technology status in Indonesia and a qualitative asssessment of integral reactor concept.

INTRODUCTION

In Indonesia, a developing country with low per capita gross national product, energy consumption since 1970 has been continually increasing in support of the development in all sectors. In case of electricity, the rate of increase of electricity consumption during the last two years has been more than 15% per year. This poses serious challenges to both natural and financial resources of the country. The fossil energy is not unlimited. Hence, it is necessary to diversify our energy resources. The integration of Nuclear Power Plants (NPPs) in energy supply development has been considered based on the national energy policy which is stipulates among others : diversification, environmental concerns and conservation in support of sustainable development.

A comprehensive and in-depth Feasibility Study has been under taken since November 1991 to give strong justification to Nuclear Power Plants introduction in our electricity system. This study covers both techno-economic and safety aspects as well as site and environmental aspects.

Based on the Feasibility Study Report, also called Non-Site Study Report, which is submitted on December 30, 1993, coal fired plants will dominate the electricity generation system. The optimization shows that the first NPP is feasible in the year 2004. The contribution of NPP will increase in accordance with the demand.

Regarding the reactor nuclear technology and safety requirements, two steps approach has been considered : for the next 50 years period and beyond 50 years. In the first period, the NPPs will be dominated by present proven system with some evolutionary improvements in technology and safety. Beyond 50 years, new evolutionary and revolutionary type of reactors may emerge, with very special characteristics in terms of conceptual, safety, fuel and performances aspects. In this period, the integrated type of reactor, which is a complete departure from current LWR design, might have a great chance to enter in the market.

Yu. I. OREKHOV, R. S. POMET’KO, Yu. A. SERGEEV Institute of Physics and Power Engineering,

Obninsk, Kuluga Region,

Russian Federation

Abstract

In addition to the evelutionary improvements and development of new additional devices and safety systems, much prominence is now given to the search of technical principals providing the enhanced safety for a NPP using new reactor design concepts as a whole. One of such concepts refers to the experience of the Institute of Physics and Power Engineering (IPPE), accumulated in the area of the exploitation of reactor channel loops for research purposes. A long-term experience of using such loops reveals that they feature a high reliability, safety and high technical and engineering parameters.

Introduction

In addition to the evelutionary improvements and development of new additional devices and safety systems, much prominence is now given to the search of technical principals providing the enhanced safety for a NPP using new reactor design concepts as a whole. One of such concepts refers to the experience of the Institute of Physics and Power Engineering (IPPE), accumulated in the area of the exploitation of reactor channel loops for research purposes. A long-term experience of using such loops reveals that they feature a high reliability, safety and high technical and engineering parameters.

There are on-going research and development activities in every one of the aspects mentioned above:

— A critical facility is being finished, and neutronic tests on the core are scheduled to

begin as soon as fuel for the facility is available. These tests are needed to validate the codes used for neutronic calculations.

— A mock-up of the protection system is being designed and will soon be under test.

— A prototype of control rod drive was tested last year. Results of these tests were

used to design and built a full scale prototype, which is now being tested at atmospheric pressure. Follow-up tests at full pressure and temperature are scheduled to begin this year.

The Republic of Korea has recently initiated design work for an integral reactor to be used for power generation and sea water desalination. The design dates from mid 1994 and the schedule points to construction around 2005. The primary vessel is contained in an outer safeguard vessel, half-filled with water, and is designed to the same pressure as the primary. Residual heat removal in emergency is through the vessel wall to the water in the safeguard vessel and from there, by heat pipe to a cooler outside the containment. The internal pressurizer uses nitrogen gas for pressurization, with pressure driven sprays and no heaters. The heat exchanger is a once through helical one, giving 30 C super heated steam There is a steam injector to drive a containment spray system. A new control rod drive mechanism(CRDM) is under development, giving finer movement than the previous Korean magnetic jack type The fuel elements are hexagonal An extensive research and development programme is envisaged

3 6 Japan

Development work at the Japanese Atomic Energy Research Institute (JAERI) has concentrated on maintenance and cost estimates of the Marine Reactor-X (MRX). This is a compact system with forced circulation of the primary, in-vessel control rods and a water filled containment, cooled by a natural circulation system The water filled containment eliminates the need for a secondary heavy shield, giving weight advantages even over a diesel system when the weight of fossil fuel to cross the Pacific Ocean is taken into account With a fleet of 20 ships, such a nuclear powered ship was

shown to be economically better compared with a diesel powered ship. The suggested mode for maintenance and refueling (every four years) is to lift out the entire core with its containment and to replace it with another one. The estimated time for this operation is three weeks. The same principle would be used for decommissioning, in an appropriate facility.

The phases accomplished for the Start-Up Program are:

• Hydraulics characterisation, cold state.

• Primary isolated: selfpresurization test, control loops calibration. Degasification of primary side

• Whole system operation. Secondary side in liquid phase. Initial calibration of condensed tank control loops in temperature and flow. Thermal balance.

• Whole system operation. Operating regime at low power with overheated steam and automatic control.

At present the following results has been obtained:

Self pressurization in the operating range of CAREM-25 with natural convection was confirmed.

The Thermalhydraulic Process was controlled during the different power conditions without major problems The production of over-heated steam is compatible with the CAREM operating conditions.

Considering the loop as an experimental machine the following experience has been obtained:

The Data Acquisition and Control System has fulfilled its required performance. However, some problems related with hardware characteristics aroused: temperature were measured with an accuracy lower than required. An interface between the DACS and the sensors was frequently in needs of recalibration. They are already replaced and a safety controller was added to the Supervision System Some improvements were made: New Fuel Assembly spacer (CAREM design) in order to make a durability test of this component; new measurements in the heat exchanger and new improvements in the valves used for control. Some modification in the condenser design were also needed to reach the full power state (300 Kw of electrical heat). In order to reduced the experimental errors several data will be also recorded through a new interfase, with a lower error and frequency of calibration needs. Improvements in the thermal isolation were also done.

In the design of ISIS emphasis has been put in the prevention of core damaging accidents.

The two main safety functions, reactor shutdown and decay heat removal, are performed without recourse to the usual sensor — logic-actuator chain, i. e. with no inputs of "intelligence", nor external power sources or moving mechanical parts, according to the definition of Category В Passive Components (ref./2/).

An active Reactor Protection System, aimed at anticipating passive system interventions, is included in the design, but is not credited in the safety analysis.

As anticipated in the Reactor System Description, mixing of the Primary Water with the Intermediate Water and the consequent natural heat transfer toward the Reactor Pool is the basic feature to assure safety under Design Basis Accidents such as Loss Of the Station Service Power and Loss of Heat Sink (ref. /3/).

During these DB Accidents the pressure boundary integrity assures the availability of water to cool the core and to transfer the decay heat to the Reactor Pool.

In case of LOCA Accidents, the Core shutdown and cooling functions are possible only if a sufficient inventory of water remains available.

The design features of ISIS guarantee the availability of this water because of the prompt self-depressurization of the system which is the consequence of the same hot-cold water mixing process.

To illustrate the effectiveness of this self­depressurization capability, the two following DB Accidents are presented :

— double ended break of the lower pipe

connection between RPV and Pressurizer;

— Steam Generator tube rupture.

Additionally, the extremely fast transient

following an hypothetical break at the bottom of the RPV is reported as an exercise to better understand the therm alhydraulic phenomena linked to the self-depressurization.

All transient analyses have been carried out using the RELAP5 computer code, with a nodalization made up of 256 control volumes 262 flow junctions and 78 heat structures; neutronic point kinetics has been used to evaluate the power in the core.

Loss Of Coolant Accident

This accident consists in a double ended break of the lower, 150 mm nominal diameter line connecting the RPV and the Pressurizer. This accident scenario has been chosen because this is

the largest line of the pressure boundary and also because the break location is far from both Density Locks, thus worsening the loss of cold water from the vessels (ref. /4/).

Considering that, the break location is 25 m below the Reactor Pool water level, the absolute pressure at the break outside the RPV is 3.5 bar. No action is credited of any active protection or control system.

When the accident starts, interconnected thermalhydraulic phenomena occur simultaneously within both RPV and Pressurizer. Cold water outflows from both RPV and Pressurizer; hot primary water replaces the losses in the Intermediate Plenum through both Density Locks. This phase lasts about 2-3 seconds. Then flashing hot water causes Primary Pumps cavitation which, in turn, allows the inlet of intermediate water into the primary system and the Core via the Lower Density Lock with a quick decrease of generated power.

The Reactor behaviour can now be explained considering that the Primary Pumps remain cavitating all over the transient and the primary system behaves like two channels hydraulically connected in parallel.

Both channels, the one made up by the Core and the Riser, and the second by the Downcomer and the SGU, are alternatively flooded by intermediate water entering the primary system through the Lower Density Lock. Self-depressurization of the system takes place mainly because of the following two water mixing effects (fig. 3):

In the RPV. hot primary water flowing from the Upper Density Lock mixes up with the large volume of cold intermediate water of the RPV Head, purposely provided for this function 1.

In the Pressurizer. hot water flowing down through the vertical pipes mixes up with the large volume of cold intermediate water underneath.

The system pressure at the break equals the external pressure in about 450 seconds.

1 Thermalhydraulic phenomena in the large plenum of the RPV Head would be better predicted by a 3D code which is, in any case, needed to optimize the design of the internal structures to enhance water mixing in this region.

At this moment the RPV water stops flowing out and reversal flow of cold, high-boron water from the Reactor Pool sets on.

The core is shutdown (fig. 4) by intermediate water entering through the Lower Density Lock.

Later on in the transient, reversal flow from the Reactor Pool starts recovering the water level in the RPV; at the end of the computer run (i. e. after 900 seconds) about 40 t of water have already entered the RPV from the Reactor Pool.

During the transient the Core never uncovers or heats up as shown in Figure 6. The maximum temperature of the "average" fuel rod has remained lower than at nominal conditions. A similar behaviour is shown for the clad surface temperature.

Steam Generator Tube Rupture

In this accident a break of 10 cm^ cross section located at the connection between SGU
tubes and steam headers is simulated; the break size is approximately equivalent to the cumulated cross sections of 8 SGU tubes.

No credit has been taken for action of active systems that can mitigate the consequence of the accident, but for the Primary Pumps Speed Control System which delays the inlet of highly borated water through the Lower Density Lock. The steam pressure and the feedwater flow rate are assumed accordingly to remain constant during the transient.2

When the accident occurs, water from the primary system enters the SGU ruptured tubes at a max mass flow rate of 96.5 Kg/s.

An equal amount of intermediate water enters the primary system through the Upper Density Lock as long as the Primary Pump Control System is capable to control the hot-cold interface level in the Lower Density Lock.

Primary water with increasing boron concentration enters the core and reduces the generated power (fig. 7).

transient.

Both effects of reduced core power with associated lower primary water temperature and Pressurizer self-depressurization reduce the overall primary system pressure (fig. 8) down to the secondary system pressure (tube-side SGU pressure) which has been assumed to remain at its nominal value.

At this time the primary water stops flowing into the SGU tubes. Figure 9 shows that the cumulated amount of water loss is less than 8 tonnes which corresponds to the inventory of the hot water in the pressurizer.

The curve of the fuel temperature shows a steadely decreasing pattern, fig. 10.

Break at the bottom of the Pressure Vessel

In this exercise an hypothetical break of 500 cm2 cross section has been assumed to occur at the bottom of reactor pressure vessel; this accident scenario is arbitrary and imagined to generate a very severe thermalhydraulic transient; in fact the break is positioned at the lowest location of pressure boundary and therefore has the potential of completely emptying the RPV. This exercise is intended to demonstrate that the self-depressurization process can avoid the uncovering of the core even in this case. No protection or control systems, no any other active system was credited during the accident analysis.

When the transient starts, there is a large blowdown of intermediate water from the RPV and Pressurizer into the Reactor Pool (the initial mass flow rate through the break is about 7000 kg/s). The escaping flow rate is fed by displaced primary water which is mostly contained in the SGU. Primary water leaves the SGU from the bottom via the Downcomer and, after few seconds, also from the top via the Primary Pumps and Upper Density Lock.

At the very begining of the transient the water flowing down through the Down earner splits in two streams: the one leaves the Inner Vessel through the Lower Density Lock and the second flows up through the Core, the Riser and leaves the Inner Vessel through the Upper Density Lock. The reactor core is continously fed

by primary water flowing upwards and its temperature is continuously decreasing because it is kept cooled since the beginning of the transient.

At the time of about 7 seconds, with Primary Pumps in cavitation, the primary water stops leaving the Inner Vessel through the Lower Density Lock and a reversal flow of intermediate water sets on flooding the core.

At this moment the usual way of natural circulation of ISIS reactor is recovered and the primary system fed with cold and borated water.

The mixing of cold and hot water initiates the self-depressurization of the system in the way described before for the case of LOCA (fig. 11).

The system continues its depressurization up to the time of about 200 seconds when its pressure drops below the Reactor Pool pressure at the break location.

The evolution of the generated power is shown in fig. 13; the power reduction during the first 7 seconds is caused by the void effect associated to the depressurization and the following shutdown is assured by the borated water.

800 600 400 200 0.0

At this moment, the total mass of displaced water (figure 12) is less than 200 tonnes (approximately 50% of the total inventory of one module) and the RPV has been emptied only down to about the center line of the SGU.

After about 1000 seconds, the initial water inventory is completely recovered and the reactor is in the state of stable cold shutdown.

Fig. 15 — Break at the bottom of the RPV
Clad surface temperature at different elevations