With the natural circulation of water in the primary circuit, the flow rate increases as the power rises. There was a danger that this circumstance will restrict the rate of the MM power increase due to a slow increase of the flow rate to a new (increased) value or initiate the circulation instability.

Experiments have eliminated this possibility. In the case of a momentary raise of the power from initial values of 100, 300 and 550 kW by 200 kW, the set-up of a new flow rate value occures smoothly and readily (for < 1 min). Thus, the natural circulation is not a factor which could hinder providing the reactor with the necessary manoeuvre characteristics in the process of its operation as a part of a District Heat Plant.

2.2. Accident Regimes

For an integrated reactor, in service inspection is somewhat more complicated than for non-integrated ones. This problem arises from the fact that integrated reactors have major components included in the pressure vessel; hence, it is difficult to access to these components for inspection. The components included in the pressure vessel also make more complicated the inspection of the pressure vessel itself, because they can render inaccessible some areas of the pressure vessel.

The standards that were studied regarding in service inspection are ASME Code, Section XI and IAEA Safety Series 50-SG-O2. Most of the requirements of these standards can be fulfilled, taking in account that most in service inspection activities can be carried out during annual reactor refueling.

The problems identified to this date are:

— Inspection of welds of the pressure vessel at core level: The plate that separates the

hot leg from the cold leg prevents accessibility to these welds from the inside of the vessel. They will be inspected from the outside of the vessel.

— Inspection of longitudinal welds of the shell of steam generators: This weld can be

avoided using a complete forge for the shell. If engineering and/or manufacturing reasons call for the use of a rolled plate, there will be only one longitudinal weld, and Osteam generators will be designed so as to have this weld accessible for inspection from the inner part of the pressure vessel.

— Inspection of circunferential welds between steam generator shells and plate tubes:

Parts of these welds are inaccessible. A sampling inspection is permitted for this type of welds.

— Inspection of steam generator tubes: The top of straight tubes is readily accessible when the pressure vessel is open, thus no problem with the inspection of tubes is foreseen. To plug a tube, it is necessary to plug both ends of the tube, the lower end being accessible only through the tube. There are technics developed for plugging tubes in this geometry: it is easier than the sleeving techniques that are currently being used. A complete tube rupture could turn repair impossible: in this highly improbable event, the full steam generator can be replaced, not a very complicated operation in a CAREM.

1 Offices/administration

2 Secondary Control Room

3 Main Control Room

4 Cold/hot Dressing Room

5 Spent Fuel Pool

6 Containment

7 Waste Treatment Plant

8 Hot Laboratory

9 Decontamination

10 Hot Workshop

11 Ventilation Extraction

12 Reactor Auxiliary Systems

13 Demineralized Water System

14 Ventilation Injection

15 Fuel And Chemical Store Area

16 Neutralization Pool

17 Auxiliary boiler

18 Chillers

19 Chimney

20 Mechanical Workshop

21 Electrical Workshop

22 Turbine Hall

23 Fire And Plant Water Tank

24 Water Treatment And Pump Hall

25 Batteries/Electric Board Room

26 Transformers

4.5.1. Control mechanisms

Four types of control for integral reactor were identified in addition to soluble boron as given in the table below:

Figure 2 shows the general arrangement of the reactor pressure vessel and its internal structures. The core is low down in the vessel. Helically coiled once through steam generator is located between the core support barrel(CSB) and reactor vessel. Thermal shields are provided below the steam generator surrounding the core to reduce the neutron fluence level on the reactor vessel. There are four main coolant pumps installed on the reactor vessel above the steam generators. The upper plenum of the vessel forms a pressurizer to maintain the operating pressure.

MS/FWS	CCS

The primary coolant flows up through the core and CSB riser, through the pumps, down through the steam generator and back to the lower plenum under the core. The current design is capable of 50% of full power operation only with natural circulation.

The UNITHERM NPP thermal-hydraulic cycle (Fig. 1) includes three interrelated process loops, the last of which accommodates all heat consumers (turbine-generator unit and heat system or process steam boilers).

Selection of coolant parameters for the above-mentioned process loops was based on the proven range of operating pressures and temperatures typical for NPP water-cooled reactor primary circuits, and on the experience with mobile NPP operating with primary coolant natural circulation Besides, variation limits of the primary coolant parameters without reactivity compensation of nuclear fuel bumup by control rods were taken into account On the other hand, selection of coolant operating temperatures and pressures for the intermediate and steam-turbine •loops was greatly influenced by the conditions which could provide acceptable efficiency and orientation on the development and operating experience with steam turbine units that might be considered as existing prototypes. The analysis of thermal characteristics of steam turbine units of such type enabled selection, taking into account the above reasons, of coolant parameters for the process loop of heat consumers

Transport of heat from the primary circuit to heat consumers during the phase change in the intermediate loops reduces the required coolant flows and increases pressure m the natural circulation system, while the desire of optimal distribution of the available temperature difference between the coolants of the primary and third circuits determines the intermediate loop coolant parameters Considering the above reasons, the parameters of the UNITHERM NPP process loop coolants are as in Table 1 below.

The proposed UNITHERM NPP has been designed to employ integrated water-cooled NSSS as a heat source (Fig. 2). NSSS combines in one vessel the main primary circuit components — core, steam generator (SG), pressurizer, control and protection elements. This allows to avoid primary circuit pipework, reach extremely compact location of ionizing radiation sources and potentially dangerous working fluid — primary coolant. The NSSS design ensures core cooling and heat supply to steam generator by convection of primary coolant and thereby allows to eliminate forced circulation. Such approach to the design of the NPP main component — nuclear reactor — is necessary to reach maximum possible reliability and simplicity due to absence of active elements with continuously moving mechanical parts. The group of absorbers with

associated drives, which perform the function of emergency protection and compensation for reactivity variation, is a single movable element. The absorbers are displaced once during NPP continuous operation when starting NSSS unit in normal operation. In emergency situations it is possible to drop the absorbers which perform the function of emergency protection.

The integrated NSSS vessel made of 15X2MFA-A steel with corrosion-resistant cladding consists of the shell with welded elliptical bottom and flange.

The central part of the vessel accommodates the removable shield with the core, lattices of absorber rods and devices of additional emergency protection. The thermal shield which also functions as core reflector and radiation and thermal shield, is arranged around the removable shield in the bottom section of the vessel.

Inside the removable shield above the core there is a cavity with pressurizer equipped with a set of flat screens serving as upper radiation shield.

In the annular space between vessel and removable shield the steam generator tube system banks are located. Intermediate circuit heat exchanger is heat exchanger with coil-type heat transfer surface, where primary coolant moves in the tubes while the intermediate circuit fluid moves in the intertube space. The heat transfer surface is made up by 24 banks of the same type, the shells of which are designed to withstand the primary circuit pressure. The banks are pairwise combined into 12 assemblies which are connected to 12 sections of the intermediate circuit heat exchanger.

Steam collecting headers and water distribution chambers of intermediate circuit heat exchanger sections are located in the vessel flange area. The chambers are also connection points of SG sections to the NSSS vessel. The shell of each SG section is a cylindrical vessel with welded spherical bottom with a coil-type tube system inside cooled by the coolant from the heat consumer circuit. In each section of the SG above the tube system the coil-type system of independent cooldown circuit is arranged. From the top the shell is closed with a spherical head.

In the choice of characteristics and structure of the core and its control elements the following priority concepts have been adopted:

• maximum possible reduction of operative reactivity margin, in particular, the fraction of total efficiency of shim elements, per group of absorber rods with individual drive;

• optimal power, coolant temperature and fuel feedback factors;

• specific power density (about 15 kW/І) which guarantees specified long-term operation without fuel element clad leaking, and minimum specific levels of residual heat release for reliable heat removal in severe accidents;

• increased reliability of emergency core chain reaction suppression system by using in the control system of additional passive emergency protection channels with operation mechanism differing from that of main functional components.

To reduce core overall reactivity the adopted design philosophy excludes withdrawal of control elements from the core in power generation mode during continuous NPP operation. In this period, the core operates in self-control mode due to variation of coolant average temperature and absorber bumup.

The additional emergency protection actuator is a structure with leaktight vessel, which by its lower part enters the core instead of one fuel assembly, while the upper part is flange connected to the NSSS vessel head. Inside the actuator vessel, actuating element consisting of the accumulator with absorber and interconnected receiving chamber made of two elastic membranes. The space between the vessel and the actuating element is filled with nitrogen, control fluid. Gas (He-З or boron trifluoride) is used as absorber.

When there is no emergency signal, the membranes of the receiving chamber are under control fluid pressure and the absorber is displaced into the accumulator. The emergency signal activates the electromagnetic switch and control fluid is discharged to the NPP containment volume. By its pressure in the accumulator the absorber is displaced to the receiving chamber spacing the membranes apart. To return additional emergency protection to initial position, the control fluid from the tank outside NSSS vessel is fed to the actuator which compresses the membranes and displaces the absorber to the accumulator.

The steam generator for the VPBER-600 reactor has been developed using experience of development, experimental investigations, fabrication and operation of prototype cassette steam generators.

Beginning from the middle of the 1970s, OKBM has carried out investigations to create cassette-type one-through SGs from straight tube steam generating elements. Work has been performed on the following:

1) theoretical investigations and design developments;

2) experimental studies (development work);

3) technological development of SG units and elements;

4) creation of an up-to-date base for SG fabrication;

5) account of SG operation experience.

When designing cassette SGs all normal and emergency modes of operation were analyzed, the factors exerting the most damaging effect on the SG structure were determined, and the most loaded parts of the SG were revealed.

So, to verify SG characteristics an R&D package has been defined and performed. It includes the following experimental investigations:

1) thermotechnical and hydraulic tests;

2) hydrodynamic investigations;

3) aerodynamic investigations;

4) mechanical tests;

5) thermocyclic tests;

6) hydrocyclic tests;

7) wear resistance tests;

8) vibration tests;

9) shock resistance tests;

10) studies of stability of throttle hydraulic characteristics;

11) studies of SG operability on emergency degradation of feedwater quality;

12) SG field inspections;

13) verification of SG maintainability properties.

All the scope of the R&D work has been performed mainly on OKBM experimental facilities. OKBM has reports with test results for each test type.

OKBM has the universal experimental base allowing it to solve the entire complex of tasks arising when designing SGs. It is evident that each specific SG design should be subjected to acceptance tests aiming at demonstration of the SG characteristics to the Customer. The OKBM experimental base includes test facilities of various powers including those of 45 MW.

Mass production of the cassette-type SGs is established in Russian enterprises which have modern technological equipment including automated complexes.

Radioactive equipment and structures of AST and VPBER plants can be divided in contaminations groups according to their accumulated activity (m coirespondence with national radiological safety norms, that allows evaluation of mass, volume and methods of handling (Fig-s 7, 8).

The hot equipment and structure mass (in-vessel barrel, guide rubes-connecting device unit, etc.) is -60 t for AST and -90 t for VPBER, the medium equipment mass (reactor vessel, heat exchange equipment, etc.) is -270 t for AST and -1850 t for VPBER, the cold and clean equipment mass (guard vessel, concrete reactor silo structures, etc.) is -1000 t for AST and -3750 t for VPBER.

Liquid radwaste formed during cleaning of the circuit water, the reactor and equipment decontamination (liquid radwaste activity does not exceed 1000 Ci) should be reprocessed — it is evaporated, concentrated and bituminized (AST-500) or cementated (VPBER-600) to confine the activity.

Solid radwaste is also reprocessed it is ground, burnt, pressed and sintered It substantially decreases the total volume of radwaste

Protective cask

Fuel assembly and internals reloading

Equi pment parting mani pulator

Lifting-and-

Cuttei

Fia. 6

VPBER-600 MAIN EQUIPMENT MASS AND

RADIOACTIVITY

AST-500 MAIN EQUIPMENT MASS AND
RAPT О ACTIVITY

VPBER-600 AND WER-440 fNPP’ LOVISA ) REAC-
TOR UNIT DECOMMISSIONING CONDITIONS

A thick water layer between the core and reactor vessel in integral reactors ensures low radioactivity of the reactor structures Thus, together with the absence of a branched pipe system in the primary circuit, it reduces the quantity of radioactive wastes and simplifies decommissioning

5 2 Decommissioning concept

The low radioactivity of an integral reactor vessel can lead to a preference for "immediate dismantling" (after preparatory work) The concept of "delayed dismantling" can be still used to further reduce radiation dose to workers

For a nuclear distnet heating reactor it has to be located near the user due to the medium >f heat transmission (hot water or low pressure steam). It means a NHR is surrounded by a populous area Using emergency actions as an essential element in the ultimate protection of the public can thus become impractical Therefore, in all credible accidents the radioactive release from heatmg reactor has to be reduced to such low levels that off-site emergency actions, including sheltering, evacuation, relocation and field decontamination will not be necessary In the other hand there is a senous challenge to the economy for a NHR The capacity of a NHR can not be as big as that of NPP due to the limitation of heat transmission

Moreover the load factor is also much lower than that of NPP. It is obvious that to meet the safety requirement and lower the capital investment are the major concern in the design of a NHR. The only solution is to have a design with inherent safety characteristics and passive safety as much as possible instead of the complex engineering safety features.

The reactor structure and core cross section of NHR-200 are shown in Fig. l and Fig.2 respectively. The simplified schematic diagram is given in Fig.3. The main design data of NHR-200 arc listed in Table 1.

The majoT features of the NHR-200 design are

(1) Integrated arrangement, self-pressunzea performance and dual vessel structure

(2) Natural circulation for primary loop

(3) Passive safety’ systems including Residual Heat Removal System and Boft Injection System

(4) Low operating parameters with large safety margin including temperature, pressure and power density

With these features, the probability of pressure boundary break is much lower than that of conventional NPP The mitigation of a LOCA is much easier and there is no ECCS m the NHR The supporting systems such as on-site diesel generators, component cooling system, service water system, instrument air system and ventilation svstem etc do not provide with safety functions, namely they are non safety-related m the NHR design While in a NPP these systems are safety-related and sometimes loss of them are dominant initiators resulting in core damage

A. YAMAJI, J. SHIMAZAKI, M. OCHIAI, T. HOSHI Japan Atomic Energy Research Institute,

Tokai-mura, Naka-gun, Ibaraki-ken,

Japan

Abstract

An advanced marine reactor, MRX, has been designed to be more compact and lightweight with enhanced safety. The reactor is an integral PWR with a water — filled containment vessel, in-vessel type control rod drive mechanisms and an emergency decay heat removal system using natural convection. These are adopted to satisfy the essential requirements for the next generation of marine reactors, namely, compact, light, highly safe and easily operated. The engineered safety is accomplished through a simplified system using water in the containment vessel. The LOCA analysis shows that the core flooding is maintained even when taking into account the ship inclination. To shorten the time of the maintenance and refueling works, a one-piece removal method is proposed. This method involves removing the containment vessel with its internals and replacing it in another one whose maintenance and refueling have already been completed. The economic evaluation of nuclear ships equipped with MRXs shows that some types of nuclear container ships will hold an economically dominant position over diesel ships in the near future, because of the environmental costs of diesel ships. R&D works have been making progress in the safety study of the thermal hydraulic phenomena, in-vessel type control rod drive mechanisms, the automatic control system, the nuclear ship simulator, etc.

1. Introduction

Compared to ordinary ships, nuclear ships are capable of long and continuous periods of voyage using high power without refueling. This advantage would contribute greatly to the diversification and expansion of sea transportation and ocean development in the future, as well as contribute to the survey and research activities on a global scale, especially in the Polar region. Another advantage is that no discharge of C02, NOx and SOx occurs during navigation, which helps to prevent environmental disruption due to NOx, SOx and the greenhouse effect due to C02. Especially in regards to NOx and SOx, the discharge quantity will be severely restricted even for ships in the
future. The above mentioned characteristics strongly motivate us to develop marine reactors as an economical power source gentle to the natural environment.

On the other hand, the marine reactor should be compact and lightweight since it has to be installed in a narrow and limited space on a ship. Furthermore, a smaller marine reactor is more economical for a ship’15. It must have high safety characteristics as well as easy operation and maintenance. It should also be possible to operate automatically to a great extent, since the operation must occur in the ocean environment without aid from land.

The Japan Atomic Energy Research Institute (JAERI) is conducting a design study on an advanced marine reactor, MRX, to obtain a more compact and lightweight reactor with enhanced safety’25-’45.

The MRX is designed to use lOOMWth for a reactor plant of an icebreaker scientific observation ship, but the concept could be applied to those of general commercial nuclear ships with wide output ranges of 50 through 300MWth

Containment vessel (Inner D: 7.3m Inner H :13.0m)	Containment water cooler (4 trains) (Heat pipe type)
Control rod drive mechanism (X13) Water spray header Pressurizer heater
Perforated plate (to enhance steam condensing & water stabilizing) Emergency decay heat removal system (x3) Plate for water stabilizing (x 8)
Main coolant pump (x2) Pressure relief valve (x3) Steam generator (Once-through helical coil type: 2 trains) Reactor vessel nner D: 3.7m) Core Fuel assembly (x 19) Flow screen
Thermal insulator-
Fig.1 Concept of MRX plant

Table 1

Reactor type : Integral PWR

Thermal power (MWt) : 100

1. Core and reactivity control

Fuel/moderator material : UO2/H2O

Fuel inventory (tons of heavy metal) : 6.326

Average core power density (kW/liter) : 41

Average/maximum linear power (kW/m) : 7.626/30

Average discharge burnup (MWd/t) : 22,600

Enrichment (initial and reloaded) : 4.3/2.5%

Life of fuel assembly (year)

Refueling frequency (year)

Fraction of core withdrawn (%)

Active core height (cm)

Equivalent core diameter (cm)

Number of fuel assemblies Number of fuel rods per assembly Rod array in assembly Pitch of assemblies/fuel rods (mm)

Clad material Clad thickness (mm)

Type of control rod Number of rod clusters Number of control rods per assembly Neutron absorber material Additional shutdown system Burnable poison material : Fuel rod with Gd2(>3 and burnable

poison rod of borosilicate glass

design description

2. Reactor coolant system

(1) Coolant

Coolant medium and inventory : H2O (411)

Coolant mass flow through core (kg/s) : 1,250 Cooling mode : Forced

Operating coolant pressure(MPa) : 12

Core inlet/outlet temperature(°C) : 282.5/297.5

(2) Reactor pressure vessel

Inside diameter/Overall length (m) : 3.7/10.1

Average vessel thickness (mm) : 150

Design Pressure (MPa) : 13.7

(3) Steam generator

Number of SG : 1 (2 trains)

Type : Once-through helical coil

Configuration : Vertical

Tube material : Incoloy 800

Heat transfer surface per SG (m2) : 754

Steam/feed water temperature (°С) : 289/185

Steam/feed water pressure (MPa) : 4/5.8

(4) Main coolant pumps

Number of cooling pumps : 2

Type : Horizontal axial flow canned motor

Pump mass flow rate (kg/s) : 640

Pump design rated head (m) : 12

Pump nominal power (kW) : 145

3. Containment

Type : Water filled (simple wall)

Inner diameter/height (m) : 7.3/13

Design pressure (MPa) : 4

Design temperature (°С) : 200

depending on the type, size and velocity of ships. In addition to the icebreaker, the MRX series reactors will be favorably applied to high speed merchant ships, very large container carriers and super high-speed container ships, which need high power and long distance voyage. A view of MRX is shown in Fig. 1.

2. Design Features of the MRX

The improvement of the economy of reactor system depends strongly on the reduction of construction and operation costs. The construction cost can be reduced by means of making the compact, light and simple reactor system. The MRX adopts the following design features to be realized the above mentioned reactor system with highly passive safety : (a) Integral PWR. (b) In-vessel type control rod drive mechanisms, (c) Water-filled containment vessel, (d) Emergency decay heat removal system using natural convection. Table 1 shows the major specifications of the MRX.

(1) Integral PWR

Integral PWR could eliminate the possibility of large scale pipe rupture accidents, simplify the safety systems and reduce the dimensions of the reactor plant.

(2) Reactor core and reactor pressure vessel (RPV)

The core consists of 19 hexagonal fuel assemblies. The hexagonal assemblies, rather than rectangular ones, have been selected for reducing neutron leakage from the core and to operate with a small number of control rod clusters. The design conditions of the core specific to the marine reactor are as follows: (a) To maintain non-criticality (keff<0.99) under the

condition of normal temperature without use of a soluble poison even if one of the control rod clusters which has the largest reactivity worth is withdrawn from the core, (b) To operate the reactor with a sufficient power level for steerageway (^30% of full power) even in the case that one of the control rod clusters which has the largest worth cannot be withdrawn from the core, (c) To keep enough residual reactivity (^2%) for overriding Xe poisoning at the EOL. (d) The life time of the fuel assembly is 8 years with the plant factor of 50% (~23,OOOMWd/t). (e) The refueling frequency of 4 years with 52.6% of the fraction of the core withdrawn.

The fuel handling system is installed in land facilities. The average power density is sufficiently low (41kW/l) which shows that the core has enough margin for thermal reliability.

The RPV is relatively larger in size because of an integral PWR. This provides a larger primary water inventory with increasing the distance between the reactor core and the RPV, and reduces the neutron fluence at the RPV. The calculated value of the irradiation of fast neutrons is below 8xl0isn/cm2 (E^l. llMeV) at the inner-surface of the RPV for full power reactor operation of 20 continuous years.

(3) Control rod drive mechanisms (CRDMs)

The CRDMs are placed in the upper region inside the RPV to enhance the reactor safety with eliminating the "Rod Ejection Accident" and to achieve a compact reactor plant.

(4) Steam generator and primary circuit

The steam generator of once-through helical coil type is placed in the RPV diagonally upper region of the core. Two trains are adopted for the main steam and feed water lines. The vertical distance between the upper surface of the core and the bottom of the steam generator is selected to be 36cm to obtain a compact RPV, and an iron shield is installed between them to satisfy the design condition of the dose rate equivalent in the engine room.

The whole primary circuit is almost incorporated in the RPV except main coolant pumps, a volume control system and a residual heat removal system. The two main coolant pumps are placed in the hot leg at the upper cylindrical region of the RPV. The pressurizer is installed in the upper part of the RPV. For the maintenance and inspection, it is so designed that the reactor components in the RPV and the primary coolant pumps can be removed remotely and the steam generator tubes can be inspected from outside the RPV.