4.71. Localized stress distributions, including those at welding sites, and their effects on the mechanical performance of structures, including leak rates, should be evaluated.

4.72. For prestressed concrete containments, particular attention should be paid to identifying areas of low prestressing (such as areas surrounding large penetrations and transition zones between cylinder and basemat), so that measures can be taken to avoid fractures and leakage due to concrete creep and shrinkage. In these critical areas, if the containment has no internal liner, leaktightness should be ensured by means of a local coating, local injection of sealing products or other appropriate methods.

4.73. For containments provided with a metallic liner, the zones of anchorage of the liner to the concrete and the connections of the liner to other metallic structures such as penetrations are also critical areas. Local effects of stress in these zones should be analysed and taken into account.

4.74. The assessment of the susceptibility of structures to fatigue should be made on the basis of a complete evaluation of the stresses and cycling, including pressure cycling for testing, temperature cycling and pipe reactions.

4.206. Soft sealing materials are commonly used in multiple containment applications, such as in the sealing of ventilation valves or the inflatable sealing of air locks. Although these materials contribute to a very high leaktightness of the containment under normal conditions, their behaviour in design basis accidents should be properly demonstrated. Potentially damaging effects for soft sealing materials include embrittlement and cracking due to high tempera­tures and irradiation, dissolution due to moisture and steam, and swelling or shrinkage due to temperature fluctuations. Specific consideration should be given to the protection of these materials from the direct effects of hydrogen burning and/or the accumulation of radioactive aerosols. In extreme conditions such materials may degrade to the extent that their mechanical properties are altered.

4.207. The anticipated lifetimes of soft sealing materials and the ageing mechanisms that affect their performance should be assessed, and appropriate replacement intervals should be established (para. 4.39). Sealing components should be designed to be easily inspectable and replaceable.

I—11. The bubbling condenser containment system (Fig. I-4) in pressurized water reactors uses a concept for the suppression pool in which the high pressure steam-air mixture resulting from the conditions following a LOCA is directed through submerged tubes into pools of water. The steam is condensed in the bubbling condenser pools.

FIG. I-4. Schematic diagram of a bubbling condenser containment system for a pressu­rized water reactor: 1, containment; 2, upper containment volume (wet well); 3, lower containment volume (dry well); 4, bubbling condenser system (suppression pool); 5, suppression pool cooling system (not required if the heat capacity of the condenser system (4) is sufficiently large); 6, passive spray system; 7, active spray system; 8, filtered air discharge system; 9, liner.

I—12. The containment is a cylindrical concrete structure divided into three isolated volumes: the lower volume (dry well), which contains all the major components of the primary reactor coolant system, the bubbling condensers (suppression pools) and the main upper containment volume (wet well). Non­condensable gases (including noble gas fission products) that are driven into the bubbling condenser chambers are vented through openings into the main
upper containment volume. Radioiodine and other soluble or particulate fission products are trapped in the bubbling condenser water pools.

I—13. Open tanks located in the upper containment volume are connected through U tubes to water spray nozzles in the lower containment volume. During fast pressure transients in the containment system, the passive sprinkler system is activated by the pressure differences between the water inlet of the U tubes submerged in the tanks and the nozzle outlet. An active spray system, with an independent stored water supply, is used to provide the functions of both energy management and radionuclide management. When the water supply in the spray tanks is exhausted, a recirculation mode is initiated and water from the building sump is pumped through a heat exchanger and sprayed into the lower containment volume. After a few minutes, the pressure in the lower volume falls below atmospheric pressure and an inverse pressure differ­ential is created between the upper volume and the lower volume. Air is prevented from returning from the upper volume to the lower volume by hydroseals formed in the bubble tubes. Once the pressure in the lower volume has been reduced below atmospheric pressure, the leakage of radionuclides from it will cease.

1.4. Requirements for the design of containment systems are established in Section 6 of Ref. [1]. The objective of this Safety Guide is to make recommen­dations on the implementation and fulfilment of these requirements. It is expected that this publication will be used primarily for land based, stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat generating applications (such as for district heating or desalination). It is recognized that for other reactor types, including future plant systems featuring innovative developments, some of the recommen­dations may not be appropriate or may need some judgement in their interpretation.

1.5. This publication is intended for use by organizations responsible for designing, manufacturing, constructing and operating nuclear power plants, as well as by regulatory bodies.

SCOPE

1.6. This Safety Guide is mainly based on the experience derived from the design and operation of existing reactors, and it applies to the most common types of containment. It also includes some general recommendations for features that would be used in new nuclear power plants for dealing with a severe accident.

1.7. This Safety Guide addresses the functional aspects of the major containment systems for the management of energy, radionuclides and combustible gases. Particular consideration is given to the definition of the design basis for the containment systems, in particular to those aspects affecting the structural design, such as load identification and load combination.

1.8. Recommendations are provided on the tests and inspections that are necessary to ensure that the functional requirements for the containment systems can be met throughout the operating lifetime of the nuclear power plant.

1.9. Design limits and acceptance criteria, together with the system parameters that should be used to verify them, are specific to the design and to the individual State, and are therefore outside the scope of this Safety Guide. However, general recommendations are provided.

STRUCTURE

1.10. Section 2 concerns the safety functions of containment systems and their main features. Section 3 deals with the general design basis for containment systems. Section 4 provides recommendations for the design of containment systems for conditions in operational states and design basis accidents. Section 5 covers tests and inspections, and provides recommendations for commis­sioning tests and for in-service tests and inspections. Section 6 provides recom­mendations and guidance on the consideration given in the design phase to severe accidents.

4.107. The ice condenser containment is divided into three main compart­ments: a lower section, an upper section and the ice condenser chambers. After a high energy pipe rupture, a flow path from the lower compartment to the upper compartment through the ice condenser is established. When the high pressure steam-air mixture flows between the columns of borated ice, the steam condenses on the surface ice. If the flow of steam continues for an extended period of time, a complete meltdown of the ice will occur. Long term energy management should then be performed by some other means, for example by containment spray systems.

4.108. The design of the ice condenser system should be such as to ensure that:

— The rate of heat transfer from the steam to the ice columns is sufficient in all postulated accident conditions (i. e. that the ice loading is sufficient).

— The structures of the ice condensers maintain their geometry under any accident loading.

— The vent doors open reliably.

4.109. The heat transfer correlations used in the calculations for the ice condenser system should be based on representative tests.

4.110. The ice condenser should be designed to permit periodic maintenance, inspection and testing. The important features of the ice condenser that should be maintained during operation are the ice temperature, the total amount of ice, the uniformity of distribution of the ice, the adequacy of the flow passages between the ice columns and the operability of the vent doors. The long term behaviour of the containment systems should be considered in the design. In the course of an accident, air and non-condensable gases will flow into the upper compartment while the lower compartment becomes filled with steam. Thus the containment spray, if injected into the upper compartment only, will not reduce the pressure below a certain limit, which will depend on the ratio of the volumes of the compartments. If equipment is installed for direct energy management for the lower compartment, a vacuum relief system of an appropriate design to eliminate pressure differentials between the two compartments should be included.

5.13. Tests should be carried out to verify that the equipment in all containment systems is functional. Exceptions may be made if it is impracti­cable to demonstrate some operational characteristics under non-accident conditions or if such tests would have a detrimental effect on safety.

5.14. Tests should be carried out on all electrical wiring associated with the containment systems to demonstrate that there are no deviations from the design and that all connections are in accordance with the design.

III-17. Contact between molten core material and concrete in the reactor cavity will result in molten core-concrete interactions. This process involves the decomposition of concrete from core debris and can challenge the containment by various mechanisms, including the following:

(a) Pressurization as a result of the production of steam and non­condensable gases to the point of containment rupture;

(b) Transport of high temperature gases and aerosols into the containment, leading to high temperature failure of the containment seals and penetra­tions;

(c) Melt-through of the containment liner or the basemat;

(d) Melt-through of reactor support structures, leading to relocation of the reactor vessel and the tearing of containment penetrations;

(e) Production of combustible gases such as hydrogen and carbon monoxide.

Molten core-concrete interactions are affected by many factors, including the availability of water in the reactor cavity, the geometry and physical layout of the containment, the composition and amount of the core debris, the temperature of the core debris, and the type of concrete.

III—18. Potential longer term challenges to the containment involve slow releases of mass and energy, typified by the generation of decay heat and non­condensable gases. The risks associated with these specific challenges can be judged on the basis of probabilistic safety assessments and research studies on severe accidents relevant to the specific design of the plant. Generally, the effectiveness of any proposed design feature can be assessed by means of a combination of probabilistic safety assessment, best estimate models and computer codes, together with consideration of the effects of initial boundary conditions and uncertainties in the modelling.

III-19. The long term pressurization of the containment may also be affected by the availability or unavailability of containment sprays (or heat exchangers) and air coolers.

CONTRIBUTORS TO DRAFTING AND REVIEW

1.59. The layout of the containment should be defined with account taken of several factors that are dealt with in this Safety Guide and that are summarized below:

— Optimization of the location of the entire primary system, with particular attention paid to the enhancement of cooling of the core by natural circulation;

— Provision of separation between divisions of safety systems;

— Provision of the necessary space for personnel access and the monitoring, testing, control, maintenance and movement of equipment;

— Placement of the equipment and structures so as to optimize biological shielding;

— Location of penetrations in areas of the containment wall so as to ensure accessibility for inspection and testing;

— Ensuring an adequate single free volume in the upper part of the containment to improve the efficiency of the containment spray (if any);

— Ensuring an adequate free volume and adequate cooling flow paths for passively cooled containments;

— Limitation of the compartmentalization of the containment volume so as to minimize differential pressures in the event of a LOCA and to promote hydrogen mixing, thus preventing the local accumulation of hydrogen.

1.60. The lower part of the containment should be designed to facilitate the collection and identification of liquids leaked, and also the channelling of water to the sump in the event of an accident. The annulus between the primary and secondary containments should form a single volume to the extent possible, in order to maximize the mixing and dilution of any radioactive material released from the primary containment in the event of an accident.

1.61. Containment systems should be designed to have high functional reliability commensurate with the importance of the safety functions to be performed.

1.62. The functions of containment systems should be available on demand and should remain available in the long term following a postulated initiating event until the specific safety function is no longer needed. Periodic testing of the systems should be performed in order to verify that the assumptions made in the design, including the probabilistic safety assessment if applicable, about the levels of reliability and performance are justified throughout the operating lifetime of the plant.

1.63. The single failure criterion[4] is required to be applied to each safety group incorporated in the design (Ref. [1], para. 5.34). Containment systems that, in and after design basis accidents, perform safety functions for energy management, radionuclide management, containment isolation and hydrogen control should be designed according to the single failure criterion.

1.64. The containment structure and the passive fluid retaining boundaries of its appurtenances should be of sufficiently high quality (ensured, for example, by means of rigorous design requirements, proper selection of bounding postulated initiating events, conservative design margins, construction to high standards of quality, and comprehensive analysis and testing of performance) that the failure of the containment structure itself and the failure of the passive fluid retaining boundaries of its appurtenances need not be postulated.

1.65. The containment systems should, to the extent possible, be independent of process systems or other safety systems. In particular, the failures of other systems that have caused an accident should not prevent the containment from fulfilling its required safety functions during the accident.

1.66. Consideration should be given to the use of passive systems and intrinsic safety features, which may, in some cases, be more suitable than active systems and components.

1.67. The structures, systems and components of the containment systems should be qualified to perform their safety functions in the entire range of environmental conditions that might prevail during and following a design basis accident, or should otherwise be adequately protected from those environmental conditions.

1.68. Components of the containment systems that can be shown to be unaffected by the design basis accident conditions need no environmental qualification.

1.69. The environmental and seismic conditions that may prevail during and following a design basis accident, the ageing of structures, systems and components throughout the lifetime of the plant, synergistic effects, and safety margins should all be taken into consideration in the environmental qualifi­cation of the containment systems.

1.70. Environmental qualification should be carried out by means of testing, analysis and the use of expertise, or by a combination of these.

1.71. Environmental qualification should include the consideration of such factors as temperature, pressure, humidity, radiation levels, the local accumu­lation of radioactive aerosols, vibration, water spray, steam impingement, flooding and contact with chemicals. Margins and synergistic effects (in which the damage due to the superposition or combination of effects may exceed the total damage due to the effects separately) should also be considered. In cases where synergistic effects are possible, materials should be qualified for the most severe effect, or the most severe combination or sequence of effects.

1.72. Non-metallic materials, such as elastomeric seals and concrete, should be qualified for ageing on the basis of sample ageing tests, operating experience in the nuclear or non-nuclear industry, or published test data for the same or similar materials under the same qualification conditions. All ageing mechanisms that are significant and relevant in the expected conditions should be considered in the qualification. Techniques to accelerate the testing for ageing and qualification may be used, provided that there is proper justifi­cation. The same applies to the possibility of testing for separate effects rather than the superposition of effects.

1.73. For components subject to the effects of ageing by various mechanisms, a design life and, if necessary, the replacement frequency should be established. In the qualification process for such components, samples should be aged to simulate the end of their design lives before being tested under design basis accident conditions.

1.74. Components that have been used for qualification testing should generally not be used for construction purposes unless it can be shown that the conditions and methods of testing do not themselves lead to an unacceptable degradation of safety performance.

1.75. Qualification data and results should be documented as part of the design documentation.

4.152. Many containment designs include systems to recirculate water from collection points inside the containment envelope, either through heat exchangers or directly, for reinjection into the reactor vessel or into the containment spray system in an accident. Parts of these recirculation systems may be located outside the containment envelope, giving rise to a potential for leakage of radionuclides from pumps, valves or heat exchangers outside the containment envelope. Where a design of this type is used, provisions should be made to minimize the uncontrolled release of radionuclides to the environment resulting from such leakage, to test the leak rate periodically, and to detect and isolate accidental leaks by qualified means.

Control of leakage in buildings outside the containment

4.153. In buildings outside the containment, sources of radioactive material arising from leaks from the containment in accident conditions or from radioactive material stored in the buildings should also be confined.

4.154. In establishing the design basis for these buildings, consideration should be given to all possible events of both internal and external origin that could cause the release of radioactive material to the environment.

4.155. Design measures such as subdividing the buildings into compartments and ensuring adequate leaktightness should be adopted to minimize the dispersion of radioactive material inside the buildings. A filtered ventilation system should be provided to limit and control the release of radioactive material to the environment.

A.1. This appendix provides recommendations for the measurement of parameters for the containment systems, to allow diagnosis by the operator of developing deviations from normal operation; in particular, to allow detection of releases of coolant or other radioactive fluids within the containment. The operator can evaluate these parameters and take corrective actions at an early stage to prevent a minor failure from developing into a serious plant failure or even an accident condition. In addition, these measured parameters are used as inputs to the automatic containment isolation system and other reactor protection systems.

PHYSICAL PARAMETERS

A.2. Typical conditions causing deviations from normal operation include:

— Release of high temperature fluids,

— Leakage of high pressure fluids,

— Presence of radioactive gases or liquids,

— Fire,

— Mechanical failure of components.

A.3. The physical parameters that should be monitored within the containment differ in different reactor systems. Parameters that are typically monitored include:

— The temperatures of the containment atmosphere and of the fluid drains,

— The pressure in the containment building,

— The humidity in the containment building,

— The hydrogen concentration in the containment building,

— Water levels in the drains,

— Rates of fluid flow,

— Radiation levels and activity of airborne radioactive material,

— Radiochemical analysis of drain water,

— Visible abnormalities,

— Noise and vibrations,

— Fire.

A.4. The measurement sensitivities necessary to detect a developing deviation should be estimated by appropriate analytical methods.

Temperatures of the containment atmosphere and fluid drains

A.5. Both atmospheric temperatures and the temperatures of fluid drains should be measured.

(a) Atmospheric temperatures. A sufficient number of temperature sensors should be installed to measure the atmospheric temperature distribution throughout the containment building. In addition, measurements of the fluid temperatures of the containment air coolers may be used to estimate the temperature of the atmosphere. The data display should present the temperature distribution and the local trends in atmospheric temperatures and fluid temperatures.

(b) Drain temperatures. The temperatures should be measured in selected fluid drains (system drains and floor drains) in order to determine whether there is in-containment leakage from any steam system or pressurized water system. These temperature measurements should be recorded to show trends.