The upper part of the RPV is filled with the mixture of nitrogen gas and steam providing a surface in the primary circuit where liquid and vapor are maintained in equilibrium under saturated condition. The pressure of the primary system is equal to the nitrogen partial pressure plus the saturated steam pressure corresponding to the core outlet temperature. Thus the reactor operates at its own operating pressure matched with the system status.

The nitrogen gas partial pressure of 2 Mpa is chosen to maintain subcooling at the core exit in order to avoid boiling in the hot channel during transients.

The volume of gas space is large enough to prevent safety valves from opening during most severe design basis transient.

Reactor Coolant Pumps

The pumps are sealed type (i. e. glandless) canned motor pumps with added inertia to increase pump rundown time. With no shaft seals the small LOCA associated with seal failure in standard commercial designs is eliminated. With the primary water level lowered, they can be removed radially for servicing or replacement without having to remove the vessel closure head.

O. G. GRIBORIEV, M. P. LEONCHYK, D. E. SKORIKOV, V. V. CHEKUNOV Institute of Physics and Power Engineering, Obninsk, Kuluga Region,

Russian Federation

Abstract

This reactor concept is a development of the well known PIUS reactor.

The main IRIS (Integral Reactor with Inherent Safety) features consist in an integration of all the primary equipment into a pre-stressed concrete reactor vessel (PCRV), natural circulation in the primary circuit and a reactor design with free levels of the coolant and the cold borated water (linked through the gas pressurizer) instead of the upper density lock.

The large volume of the cold borated water in the PCRV provides not only passive shutdown the reactor in emergency (like it is in PIUS), but condensation of the vapor during accidents with the primary coolant boiling. The containment and safety systems may be considerably simplified.

The large scale PCRV makes possible to store inside the vessel all the burnt-up during the reactor lifetime fuel assemblies.

A thick borated water layer between the core and PCRV — walls allows decrease the residual PCRV activity upto an environmentally — acceptable level and to simplify reactor decomissioning.

The high level of safety makes possible siting of this reactor near population centres.

Introduction. The objective of this report is to illustrate conceptual advantages of the integral type PWR, called IRIS (Integral Reactor with Inherent Safety). The reactor concept is a result of development of well — known PIUS project ideas. This innovation establishes some new important properties.

Design features. IRIS design features are: integration of all the primary equipment into a pre-stressed concrete reactor vessel (PCRV); natural circulation in primary circuit; absence of the upper density lock. Principial scheme of the reactor is shown in fig. 1.

Instead of an upper density lock there are communicating vessels with free levels linked through the gas pressurizer and single lower density lock. One of these vessels is the total primary circuit, and another one is a tank with cold borated water. Hence, despite natural circulation in them, this communicating vessels system will support the pure and borated water interface in the density lock inherently.

Only a primary leakage can destroy this interface and cause a proportional flow of borated water into the primary circuit through the density lock.

To exclude the core overheating there are by-pass pipes from the primary circuit volume into the borated water tank. They are arranged above the level of primary coolant. If a coolant overheat was to happen, the level will rise up to the by-pass pipe’s location and come to the borated water tank. Same portion of cold borated water will enter into the core through the density lock.

Therefore, due to the large scale reactor vessel, we have an inherent passive safety system, which is always ready for action, does not impede the normal operation, independent of anybody.

Ya. K. KOZMENKOV, Yu. I. OREKHOV Institute of Physics and Power Engineering, Obninsk, Kuluga Region,

Russian Federation

Abstract

LOCA initiated by a guillotine break of the pressurizer surge line has been considered in the paper. The failure of two emergency core cooling system (ECCS) trains was also postulated, that turns the considered accident sequence into a beyond the design basis (BDB) class. Basic design characteristics of the ABV reactor and the containment system are presented as well as the factors of much importance to the accident progression.

SCDAP/RELAP5/MOD3.1 was used as the computer code for the simulation of reactor and containment system behavior in the course of the accident. Since a noncondensable driven pressurizing system was employed in the reactor design, the presence of dissolved nitrogen in the primary water was taken into account in calculations.

The important feature of the simulated accident is the primary system refilling with the water of pressure suppression pool driven by the pressure difference between containment system compartments.

In the manufacture of RPVs of integral reactors, existing manufacturing methods and materials can be used. The design of integral reactors which is noted for a large water layer between the cover and the vessel wall. substantially decreases the radiation damage of the vessel material during the operation.

F. M. MITENKOV, A. V. KURACHEN KOV,

V. A. MALAMUD, Yu. K. PANOV,

B. I. RUNOV, L. N. FLEROV OKB Mechanical Engineering,

Nizhny Novgorod,

Russian Federation

Abstract

Design bases of hew generation nuclear power units (nuclear power plants — NPP, nuclear co-generation plants — NCP), nuclear district heating plants — NDHP), using integral type PWRs, developed in OKBM, Nizhny Novgorod and trends of design decisions optimization are considered in this report.

The problems of diagnostics, servicing and repair of the integral reactor components in course of operation are discussed. The results of safety analysis, including the problems of severe accident localization with postulated core melting and keeping coriurn in the reactor vessel and guard vessel are presented. Information on experimental substantiation of the suggested plant design decisions is presented.

INTRODUCTION

The integral lay-out realized in boiling water reactors and BN-type reactors is a result of the search for optimum technical and economically substantiated decisions.

An analogous search process is also characteristic of the reactors of PWR type.

Investigations and developments allow the conclusion to be made that certain conditions integral reactors have considerable advantages as for mass and size in comparison with loop-type and unit-type plants.

Besides the integral lay-out, the reactor has advantages as for safety, quality of fabrication, mounting, building time and removal from operation. But the integral lay­out objectively complicates the reactor design and the problems of operational service, it causes the necessity to use highly reliable in-reactor equipment.

In the development of integral reactors especially important are specific characteristics of the heat exchanger (steam generator) built into the reactor, because the reactor vessel dimensions depend largely on heat exchange surface dimensions.

The lifetime reliability of the reactor components should be confirmed by operational experience as a part of operating reactor plants and their prototypes or by broadened complex representative tests at testing facilities in the conditions corresponding to operation conditions in the plant.

For some decades, OKBM specialists developed ship nuclear power plants and experience has been accumulated on the development of some equipment and the NPP as a whole, fabrication and experimental development of some equipment, designer supervision of the fabrication at the factories and in course of operation.

The afore-mentioned allowed the development of new generation nuclear power plants with integral reactors.

First of all is the reactor plant AST-500 for NDHP, which may be located in the vicinity of large cities.

The AST-500 reactor plant is the first in the group of the plants with integral PWRs. Its characteristics are widely known. Its main peculiarities are following: natural coolant circulation in the reactor, high safety level provided bv passive means.

The high safety level of the RP AST-500 was recognized by national technical and ecological expert examination, supervision bodies and a special commission PRE-OSART IAEA.

The main fundamental decisions of the NPP, such as integral reactor design, use of guard vessel, use of passive safety systems of various principles of operation with deep redundancy and self-actuation became the basis of the whole group of the developed plants of ATETS-200, VPBER-600 type and the others.

The main advantages of the integral design in comparison with traditional loop — type designs:

— localization of radioactive coolant in one vessel (excluding purification system);

— absence of large diameter pipelines and nozzles in the primary circuit;

— keeping the core under water level at any loss-of-tightness due to the proper choice of guard vessel volume;

— decrease of neutron fluence to the reactor vessel to the level, excluding any noticeable change of the vessel material properties, radiation embrittlement (fluence <1017n/cm2);

— higher completeness of the reactor plant important equipment, of the guard vessel at the delivery to the site and as a result increase in the quality of mounting the power units as a whole;

— reduction of NPP building time to the reducing of the installation work and simplification of construction work;

— considerable simplification of the technology and operations at NPP decommissioning and RP change for repeated use of NPP structures.

Possible negative consequences of the integral reactor design are the following:

— delivery of off-gauge heavy cargo from the factories;

— the necessity to increase considerably the rated load of mounting cranes at the

site.

Corresponding analysis and the experience of delivery of AST-500 reactors and guard vessel to the sites of Nizhny Novgorod and Voronezh confirm the feasibility of such delivery by the existing engineering means.

The reactor core is cooled by natural circulation in the range from full power operation to residual heat removal The hydraulic resistance along the primary circuit is dominant by the primary heat exchangers The "U" type tube bundles are adopted for PHES in order to give facilities for repairing The pitch of fuel elements is chosen to 13mm, a little tight than usual The flow resistance does not mcrease too much, but it is good for negative reactivity feedback There is a long riser on the core outlet to enhance the natural circulation capacity The high of the riser is about 6m Hence the average coolant velocity m core is 0 57m/s about 2 times larger than that in Although the core power density increases comparing with NHR — 5 the MDNBR is still larger than the limitation with a sufficient margin

A number of measurements and experiments have been earned out in NHR-5 and demonstrated that the capacity of natural circulation is sufficient to carry out the heat m the power range from full power operation down to residual heat removal Even m case of interruption of natural circulation m the primary circuit due to LOCA the residual heat of the core can be transmitted by vapor condensed at the uncovered tube surface of the primary heat exchangers [1]

Accidents of primary coolant loss, steam generator tube rupture, steam line break, feed water line break and total loss of electricity have been analyzed for safety evaluation of the MRX. Figure 3 shows typical results during LOCA obtained RELAP5/Mod2 calculation assuming the double ended guillotine break of 50mm dia. pipe occurred in the CV. Cooling by the EDRS, the CV water and the CffCS is taken into account, but the function of the RHRS is neglected. The maximum design values of pressure and water temperature in the CV are 4MPa and 200’C as mentioned in Table 1, and the allowable minimum water level in the RPV is 0.5m above the upper edge of the core, which is determined to keep core flooding taking into account of ship inclination and oscillation. Figure 3 shows that the maximum values of pressure and water temperature in the CV and the minimum water level in the RPV are І. ЗМРа, 140’C and 0.8m, respectively. These satisfy the design conditions sufficiently. Through these analyses, it has been proved that the passive safety features applied to the MRX have sufficient functions in the safety point of view.

J. P. ORDONEZ,

INVAP S. E.,

San Carlos de Bariloche,

Argentina

Abstract

The paper presents the design related aspects of operation and maintenance of the CAREM reactor and the principal features of its main components.

The paper covers three main topics: operational aspects, major components and

maintainability

Operational aspects

A strong negative thermal coefficient, the use of burnable poisons to compensate burnup and no use of soluble boron for reactivity control characterized reactor control.

Hydraulically driven control rod drives are fully contained in the pressure vessel.

The following research and development activities are being carried on:

A critical facility for testing main core characteristics is in final construction stage.

A full scale model of control rod drives is currently under test.

A full scale model of one steam generator will be constructed and tested.

Major components

Pressure vessel: The empty pressure vessel weights 100 tons. This fact facilitates both its manufacturing and transport. Internals and steam generators will be mounted on site. Being the reactor self-pressurized, no pressurizer is included.

Steam Generators: Twelve once-through steam generators are symmetrically placed inside the pressure vessel. Specific design aspects are discussed in the paper.

Containment: The containment is of pressure suppression type. Second shutdown system, pressure relief tank, and equipment and installation for manual reactor refueling and for handling of RPV internals, are all placed inside the containment. Provisions are also made for accommodating RPV internals during refueling and maintenance operations.

Maintainability

Lay-out: The balance of plant lay-out is conventional. For nuclear island layout, attention has been paid to the fact that the containment will not be accessible during reactor operation. This fact imposes special demands on equipment reliability, that will assure high plant availability.

In service inspection: it is currently under study. Inspection of pressure vessel welds will follow standard practices. Inspection of steam generators will be performed by conventional eddy-current techniques adapted to tube geometry. Other in-service inspections required by ASME XI, SS-50-SG-O2, and local regulatory authorities, are being evaluated, most of them being similar to the standard for non-integrated reactors.

Fuel and waste handling: manual refueling of the reactor implies changing 31 fuel elements (approx. 70 kg each) per year. Spent-fuel-pool capacity covers seven years of operation; afterwards, dry spent fuel element storage is considered.

Decommissioning: The CAREM concept does not impose specific conditions on plant decommissioning.

1. INTRODUCTION

CAREM reactor features have been described elsewhere /1/. This paper deals with detailed engineering aspects of its design, that point to important differences between the CAREM, and conventional non-integrated PWRs.

The main aspects to be discussed are:

— Operational aspects, including reactor control and control devices.

— Major components engineering: reactor vessel, steam generators and containment

vessel

— Maintainability, as related to lay-out, in-service inspection, fuel and waste handling,

and decommissioning.

The R&D status corresponding to each item is mentioned when convenient.

1. INTRODUCTION

Many developing countries are experiencing a rapid growth of their economies and an increasing need for the supply of both heat and electrical energy. The present primary energy production is based predominantly on fossil fuels, adding to the C02 burden on the environment. Nuclear power has the potential to bring about a substantial reduction in C02 releases arising from both heat and electricity generation. Although there was a rapid growth in the seventies and early eighties, in the nuclear share of total electricity generation, several factors including the impact of the Three Mile Island and Chernobyl accidents have resulted in dampening further growth. For harnessing the potential benefits of nuclear energy in meeting the future needs of both heat and electricity, it is necessary to develop cogeneration nuclear plants having the following characteristics:

• Low capital investment to reduce the financial risk.

• Construction period of 5 years or less to improve the economics of nuclear heat and power generation.

• High operational reliability through simplification of the plant systems.

• Enhanced safety features enabling siting near densely populated areas.

In response to these needs nuclear power plant designers in many countries have been developing new designs including integral reactors. Integral reactors are heat and/or power generation reactors, in which the primary coolant system components including steam generators(SGs), pressurizers and pumps are contained within the reactor pressure vessel (RPV). The current loop type power reactors have these components located outside the RPV.

The purpose of this report is to provide up-to-date technical information on the current status of the design and development of integral reactors in the Member States, based on information presented in two technical committee meetings organized by the Agency on the subject. Important aspects regarding integral reactor design and development are highlighted in the summary and the presented papers are included under the following headings:

• Development programmes and conceptual design descriptions.

• Specific systems and analyses.

• Operational, manufacturing and decommissioning aspects.

The CAREM, a small NPP Project owned by the CNEA /1/ has been developed jointly by CNEA and INVAP 12/ The concept was bom by the 80,s when the idea was presented at LIMA, Peru, dunng an IAEA conference back in 1984 At that time, the Argentinean nuclear experience was several RR built (RA-O, RA-I, RA-2, RAO, RA-6, RP-0) several Projects on RR. an enrichment plant and the experience gained in the follow up of two NPP m operation (CNA-I and CNE) It was thought that the next step to reach the nuclear maturity should be to work in the design and construction of a NPP This development in a medium size economy and restricted financial resources should have a limited nsk and so the focus was addressed to small NPP In addition the NPP should be able to operate in isolated cities, a common situation in Argentina That meant reduced ability to obtain help in operational incidcnts/accidents by the operating personnel, reduced industrial/ technical resources in the site, long distances from fully developed areas, reduced roads and transportation capacities And being the first domestic NPP, it was also considered convenient to work in an electrical plant with a small output when connected to the national electric network Similar conditions could be find in a good number of countries

The above mentioned critenum developed in the CAREM technical characteristics i e a LWR, with Integrated Primary System and extensive use of the so called passive systems (e g natural circulation cooling)

Fixed arbitranously the power in 15 Mwe, a R&D program was started and so the construction of a thermalhydrauhc lab and a critical facility

At present the work is addressed to the 25 Mwe but a new version of 100 Mwe is also foreseen

1.1. Organisation

CNEA has subcontracted the engineering of the CAREM to INVAP However by itself is working on the development of the Tuel Elements and in part of the nuclear instrumentation The follow up activities on the CAREM Project arc made by CNEA with a Coordination Group which receives all the engineering from INVAP Afterwards the engineering enter in a revision process inside the CNEA by groups specially appointed by the CAREM Coordinator Every three months the Coordination Group have a meeting with their counterparts in INVAP were technical discussions arc held Also visit and revisions of the experiment works are done In some especial topics as RPV design and manufacturing, foreign experts are invited to assist and make special revisions Experts from other companies are also sometime required