Russian Federation

I — 1. DESCRIPTION OF A NUCLEAR INSTALLATION WITH THE KLT-40S REACTOR

The KLT-40S is a modular reactor unit developed for a pilot floating nuclear cogeneration plant (PATES, in Russian), currently under construction in Severodvinsk, the Russian Federation. The KLT-40S nuclear installation belongs to a class of pressurized water reactors. The KLT-40S reactor unit is shown in Fig. I-1. Major specifications of the KLT-40S nuclear installation are given in Table I-1. A detailed design description of a floating NPP with KLT-40S reactor installations is provided in [I-1].

The main design features of the KLT-40S are the following:

— Modular design of reactor unit: the reactor, the steam generators (SGs) and the main coolant pumps

(MCPs) are connected with short nozzles, without using long pipelines;

— Four-loop reactor cooling system with forced and natural convection of the coolant in the primary circuit;

— Leaktight primary circuit with canned motor pumps and leaktight bellows type valves;

— Once-through coil type SGs;

— Gas based pressurizer system in the primary circuit;

— Use of passive safety systems;

— Use of proven techniques for equipment assembly, repair and replacement; incorporation of proven

diagnostics equipment and proven monitoring systems.

The KLT-40S core is based on marine reactor technologies and incorporates materials that are exempted from the IAEA definition of direct use material.

To increase uranium fraction, a closely packed assembly structure of the core is adopted, which provides maximum possible fuel volume in a given core volume. The core contains fuel rods with cylindrical claddings made of corrosion resistant zirconium alloy. The fuel rods are similar to those of the ice-breaker reactors but incorporate fuel with higher uranium fraction; such fuel is based on uranium dioxide granules in the inert matrix.

Each reactor unit of the floating nuclear power plant (NPP) is located in a containment that is a leaktight physical barrier designed to limit the propagation of radioactivity and to localize fission products in case of a loss of coolant accident (LOCA), using emergency containment cooling systems.

The containment is designed for internal pressure typical of design basis accidents and beyond design basis accidents, taking into account the emergency temperature conditions. The design value of the containment leakage rate ensures maximum possible limitation of the emergency planning area.

The containment, along with the barge structures, is designed for design basis external impacts including a floating NPP sink.

Protection of the systems important for safety from external impacts is provided by a protective enclosure. The protective enclosure is a waterproof and gas proof structure included in a ship hull; it covers the containment and the liquid and solid radioactive waste storage, and provides additional limitation of a leakage of radioactive products to other parts of the floating power plant and to the environment, in case of a severe accident.

The containment and the radioactive waste storage are placed in a power compartment located in the middle part of the floating power unit.

A general view of the floating power module is shown in Fig. I-2.

The floating power unit (FPU) is a flat deck non-self-propelled ship with a developed multilevel superstructure. An all-welded vessel of the floating power unit has ice reinforcements and special means for hauling and shoring. Nine waterproof bulkheads rising up to the top deck divide the FPU vessel into 10 impermeable compartments.

1- REACTOR

2- STEAM GENERATOR

3- MAIN CIRCULATION PUMP

4- CPS DRIVES

5- ECCS ACCUMULATOR

6- PRESSURIZER (1st vessel)

7- PRESSURIZER (2nd vessel)

8- STEAM LINES

9- LOCALIZING VALVES

10- HX of PURIFICATION AND COOLDOWN SYSTEM

FIG. I-1. General view of the KLT-40S nuclear installation.

The floatability of the FPU is provided in case of flooding of any two adjacent compartments for all specification load cases satisfying the requirements of the Russian Marine Register.

I-2. PASSIVE SAFETY DESIGN FEATURES OF KLT-40S

Passive safety design features of the KTL-40S nuclear installation include both inherent safety features and dedicated passive (safety) systems.

TABLE I-1. MAJOR SPECIFICATIONS OF THE KLT-40S POWER PLANT Characteristic Value Thermal power, MW 150 Primary circuit pressure, MPa 12.7 Coolant temperature, °C: — at core outlet 317 — at core inlet 279 Parameters of superheated steam downstream of the SG: — pressure, MPa 3.73 — temperature, °C. 290 Feedwater temperature, °C 170

The so-called self-protection of a nuclear installation is expressed in its capability to prevent the occurrence and to limit the propagation and consequences of initiating events which could lead to accidents. Self-protection is, inter alia, achieved by reliance on natural feedbacks and processes that require no operator intervention, no external power, and no assistance from emergency teams for a certain period of time which could be used by personnel to evaluate the situation and to undertake necessary corrective actions.

The self-protection of the KLT-40S is provided by the following features:

(a) Negative reactivity coefficients on fuel and coolant temperature and on specific volume of the coolant; negative reactivity coefficients on steam density and integral power;

(b) High thermal conductivity of the fuel composition defining its relatively low temperature and, correspondingly, low stored non-nuclear energy;

(c) Adequate level of natural circulation flow in the primary system;

(d) High heat capacity of the nuclear installation as a whole, resulting from high heat capacity of the primary coolant and metal structures, from the use of a ‘soft’ pressurizer system[28], and from a safety margin

provided for by the design for the depressurization pressure of the primary system under emergency pressure increase;

Compact design of the steam generating unit, with short nozzles between the main equipment items and with no large diameter primary pipelines;

The use of restriction devices in nozzles connecting the primary circuit systems to the reactor, which limits the outflow rate in case of a break; the location of the connection nozzles is selected so that they provide a fast transition to the steam outflow of the primary coolant in case of a break in the corresponding pipeline;

Favourable conditions for the realization of a ‘leak before break’ concept in application to structures of the primary circuit, provided by design;

The use of once-through steam generators, which limits the rate of heat removal via the secondary circuit in case of a steam line break accident.

The active and passive safety systems (see Fig. I-3) are incorporated in the design of the KLT-40S to carry out the following safety functions:

—Emergency shutdown of the reactor;

—Emergency heat removal from the primary circuit;

—Emergency core cooling;

—Localization of released radioactive products.