ML23331A900
| ML23331A900 | |
| Person / Time | |
|---|---|
| Site: | 07103103 |
| Issue date: | 08/21/1999 |
| From: | Office of Nuclear Material Safety and Safeguards, Secured Transportation Services |
| To: | Boyle R US Dept of Transportation (DOT), Office of Hazardous Materials Safety |
| Shared Package | |
| ML23331A891 | List: |
| References | |
| CAC 001794, EPID L-2023-DOT-0004 | |
| Download: ML23331A900 (24) | |
Text
R E G U L A T I O N No. 195/1999 Sb.
of the State Office for Nuclear Safety
of August 21, 1999
on Requirements on Nuclear Installations for Assurance of Nuclear Safety, Radiation Protection and Emergency Preparedness
The State Office for Nuclear Safety stipulates in accordance with § 47, par. 7 to implement § 13, par. 3, ad d) and Part A), item I.2 of the Annex to the Law No. 18/1997 Sb.,
on peaceful use of nuclear energy and ionising radiation (Atomic Law) and on amendment and supplement of several laws:
PART I
General Provisions
Subject of Regulation
§ 1
The Regulation stipulates the requirements on the design of nuclear installations for the assurance of nuclear safety, radiation protection and emergency preparedness.
§ 2
For the purposes of this Regulation it is understood
a) the lowest technically achievable values of doses of ionising radiation mean the values optimised from the viewpoint of radiation protection according to special Regulation 1),
b) the normal operation means all states and processes of planned operation of nuclear installation within the specified limits and conditions of safe operation of nuclear installation; there are namely the repeated putting the reactor into criticality, power operation and shutdown of reactor, amplification and reduction of its power, maintenance, repairs and fuel exchange,
c) the abnormal operation means all states, processes and events deviating from the normal operation, that are not planned, but which are expected to occur during the operation of nuclear installation; there are, e.g., a fast shutdown, sudden drop of load, fall-out loss of
1 turbine, loss of feed of supply of network, fall-out loss of main circulating pump etc.;
these operating stages must not result to the damage of fuel system or to the failure of fuel elements and to the failure of primary circuit integrity; after their ending, respective removing the causes and consequences, the nuclear installation is capable to perform the normal operation,
d) the fuel element means the structural component, the basic part of which is the nuclear fuel; it includes the cladding, fuel pellets, fill gas, springs, end closures etc.,
e) the fuel assembly means the set of fuel elements, that is not commonly dismantled during the fuel reloads in the reactor; it includes except of the fuel elements the spacer grids, top and bottom nozzles, further on, if they are used, the guide tubes for internal instrumentation or for bundles of regulating rods or for neutron sources, or for assembly with discrete burn-up absorbers and the shroud of fuel assembly.
f) the fuel system means the fuel assemblies and their components, the internal control components of core as reactivity control rods, burnable poison rods, if they are used, rods with neutron sources, supporting plates etc.,
g) the failure of fuel elements means the breach of fuel rod cladding pressure boundary, and consequently the possibility of leakage of the fission products into the environment,
h) the damage of fuel system means the damage of fuel element or the exceeding the dimensional tolerance for operational states or the reduction of functional capability below those assumed in the safety analyses,
i) the design limits for normal and abnormal operation mean the values of parameters, to their achievement there is assured the capability to fulfil designed functions and to prevent non-permissible leakage of radionuclides into the environment,
j) the accident conditions mean all events caused by a failure or damage of civil constructions, technological systems and components, by external events or by human errors, which lead to the violation of the limits and conditions of safe operation, and that can result to a fuel element damage or a break of fuel elements,
k) the design basis accident means an accident considered in the nuclear installation design, that may have a consequence as the leakage of radionuclides, ionising radiation or the person exposure,
2 l) the maximum design basis accident means such design basis accident considered in the nuclear installation design with the maximum radiation consequences,
m) the limiting fuel element parameters mean the maximum fuel elements parameters and degrees of their damage, which must not be exceeded under the conditions of normal and abnormal operation,
n) the single failure means an event resulting to the loss of capability of an component to perform the determined function and at the same time all the other components correctly operate; multiple failures resulting from a single failure are considered to be a part of this single failure.
PART II
Basic Requirements on Nuclear Installations to Assurance
of Nuclear Safety, Radiation Protection and Emergency Preparedness
§ 3
Defence In-Depth
The nuclear safety of nuclear installation shall be assured through the defence in-depth, based on the use of multiple physical barriers preventing the propagation of ionising radiation and radionuclides into the environment, and the multiple use of system of technical and organisational measures being used to the protection and preservation of effectiveness of these barriers, and at the same time to the protection of personnel and other persons, inhabitants and environment.
§ 4
Quality Assurance Requirements
(1) The civil structures, technological sets and components (further on only "components") important for the nuclear safety of a nuclear installation and for the radiation protection shall assure their reliable function under normal and abnormal operation conditions, and the capability to limit the consequences of failures and incidents.
(2) The components important for nuclear safety of a nuclear installation shall be designed in such a manner that they may enable to carry out during the operation the inspection of state and tests of their functional capabilities and reliabilities by methods
3 corresponding to the present state of art of a science and technology. The technical design of these components shall content the safety measures compensating (making up for) the occurrence of non-detected failures during the nuclear installation operation.
(3) The quality and suitability of computational programs used for the analyses important for nuclear safety shall be verified.
§ 5
Radiation Protection in Nuclear Installation Structures and in Environment
The nuclear installation shall have assured the radiation protection in its structures and in its environment according to the special Regulation 1).
§ 6
Emergency Preparedness On-site and Off-site of Nuclear Installation
The nuclear installation shall have assured the on-site and off-site emergency preparedness of nuclear installation according to special Regulations 2), 3).
§ 7
Protection against Component Failures
The components important for the nuclear safety shall assure under the normal and abnormal operation conditions, tests and at the origination of accident conditions, in order that their damages must not occur as a consequence of failures of other components located within the nuclear installation. Therefore, they shall be capable to withstand the changes in the environment connected with these failures, and they shall be appropriately located and adequately protected against dynamic and other impacts (missiles, pipe vibrations, liquids leakage, overloading caused by higher pressure).
§ 8
Heat Removal
(1) The technological sets and components participating in the removal of heat released as a result of nuclear fission, of residual heat and of operational heat, shall reliable assure the adequate reactor cooling under the normal and abnormal operations and under the accident conditions.
4 (2) The systems of heat removal shall be suitably redundant, physically separated and simultaneously interconnected in such a manner that they may fulfil their function also at a single failure.
§ 9
Fire Protection
(1) The components important from the viewpoint of nuclear safety shall be designed and located so that, they may fulfilled the requirements of fire protection and other requirements stipulated by special Regulations 4).
(2) At components important from the viewpoint of nuclear safety they shall be used non-combustible materials or materials with diminished combustibility.
(3) Structures of a nuclear installation shall be equipped by the electric fire signalling and by the fire fighting devices designed in such way, that in the case of failure or of random start-up, the functional capability of components important from the viewpoint of nuclear safety might not be influenced.
(4) The assessment of fire risk shall be performed for structures important from the viewpoint of nuclear safety.
(5) The nuclear installation, the part of which is a nuclear reactor with a power greater than 50 MWt, shall be assured by the fire protection brigade of licensee since the construction stage.
§ 10
Protection against Effects Caused by Natural Conditions or by Human Activity Outside of Nuclear Installation
(1) The components important for nuclear safety of a nuclear installation shall be designed in a such way, that under conditions of such natural events, which could be reasonably considered (earthquakes, windstorms, floods, etc.) or under the events resulting from the human activities outside of nuclear installation (aircraft crash, explosions in the plant vicinity etc.) it may be possible
a) to safe shut down the reactor and to keep it in sub-critical state,
b) to remove the residual power of reactor for a sufficiently long period,
5 c) to assure that appropriate radioactive leakage does not exceed the limit values stipulated by the special Regulation 1).
(2) In the design of nuclear installation it shall be considered
a) the most important natural phenomena that have been historically reported for the site and its surrounding vicinity, extrapolated with a sufficient margin for the limited accuracy (uncertainties) in values and in time,
b) the combination of the natural phenomena or effects resulting from the human activities and the accident conditions caused by these effects.
§ 11
The nuclear installation shall be designed in such a manner, that the physical protection of nuclear installation and of nuclear materials may be provided 5).
§ 12
Joint Use of Components
At a nuclear installation composed of several units, which assumes the joint use of components important for nuclear safety, it shall be demonstrated that such design shall not influence its safe operation. At the same time it is also necessary to consider the case of origination of accident conditions in one unit and the capability of correct shutdown and cooling the remaining units.
PART III
Reactor Core
§ 13
Reactor Core Design
(1) The reactor core and the associated cooling, control and protection systems shall assure with a sufficient margin that the specified design limits will not be exceeded during any operational state.
6 (2) The reactor core and the associated cooling, control and protective systems shall be designed so that in the power operating range the net effect of the prompt feedback in the core will counteract (tend to compensate for) against rapid reactivity increase.
(3) The mechanical components making up the reactor core or mechanical components located in its vicinity, including their fastenings, shall be designed in such a manner, that they withstand the static and dynamic loads under the normal and abnormal operations. At the accident conditions their possible failure must not prevent the safe reactor shutdown and the core cooling.
§ 14
Fuel System Design
(1) The fuel system shall withstands the design irradiation in the reactor core, without its damage under conditions of normal and abnormal operation despite all considered processes of deterioration of material properties and of environs conditions that can occur during the operation.
(2) The considered processes of deterioration of material properties and of conditions of environs shall include the effect of external pressure of the coolant, the increase of internal pressure in the fuel element due to fission products, irradiation of fuel and of other materials of fuel assembly, changes in pressures and temperatures resulting from power changes, chemical effects, static and dynamic loads including the coolant flow induced loads and the effect of mechanical vibrations and changes in a heat transfer, that may originate as a consequence of deformations or chemical effects. The uncertainties in dates, calculations and the manufacturing tolerance shall be considered with a corresponding margin.
(3) The specified design fuel limits for normal and abnormal operations, including the admissible fission products leakage, must not be exceeded under normal and abnormal operation, and at the same time the conditions that may occur at the core under the abnormal operation, must cause no additional significant deterioration of design properties of fuel system. The fission products leakage shall be kept under the minimum value, that is practically achievable.
(4) Under the conditions of design basis accident there the fuel elements and assemblies shall remain in their positions and they shall not suffer such damage, that might obstruct the control rods insertion or disallow effective core cooling.
7 (5) The specified design fuel limits for design basis accident conditions must not be exceeded.
(6) The structural design of fuel assemblies shall permit the adequate inspection of their parts.
(7) The design of the fuel assemblies shall be sufficiently experimentally or operationally documented.
§ 15
Neutron Flux Distribution
(1) For all levels and distributions of the neutron flux that can arise in all states of the core, including those after shutdown or during or after fuel refuelling and those arising from the abnormal operation and accident conditions, provisions of § 14, par. 1 to 5 shall be met.
(2) The means for the detection of neutron flux distribution shall be capable to detect the regions of the core in which the level and distribution of neutron flux might cause the exceeding the design fuel limits for normal and abnormal operation and for accident conditions. The core design shall support the capability of control system to maintain the level and distribution of neutron flux within stipulated limits under all core states during normal and abnormal operations.
(3) The core and associated coolant system, control and protection systems shall be designed to assure that the power oscillations which can result in exceeding the specified design fuel limits are not possible or that they shall be reliably and immediately detected and suppressed.
P ART IV
Control and Protective Systems
§ 16
Control Systems
(1) The control systems of nuclear installations shall be instrumented to monitor, measure, record and control the operational parameters important for assurance of nuclear safety under normal and abnormal operations and under the accident conditions. The communication and control devices shall be designed and located to ensure the continued
8 information flow on the operational conditions of the nuclear installation to the operating personnel and to enable to it, if necessary, a prompt intervention. The control systems shall give required signals on deviations of the important operational parameters and processes from the allowable limits.
(2) The control systems shall continuously in regular intervals or according to demand to record the values of parameters, which are according to the accident analyses important for nuclear safety of a nuclear installation.
(3) At the origination of accident conditions the instrumentation shall provide
a) the information on the immediate state of the nuclear installation, on the basis of which the protective measures may be carried out,
b) the basic information on the accident scenario and their record,
c) the information enabling to characterise the propagation of radionuclides and of radiation into the vicinity of the nuclear installation in such a way, in order that it might be timely possible to carry out the protective measures.
§ 17
Protective Systems
The nuclear installations, the part of which is a nuclear reactor shall be equipped by the protective systems, that shall be
a) capable to distinguish the abnormal conditions and automatically initiate the operation of appropriate systems, including the reactor shutdown systems according to § 21 in such a way, in order that it may be assured, that design limits may not be exceeded,
b) capable to distinguish the accident conditions and to initiate the operation of appropriate systems designed for the mitigation of consequences of such conditions,
c) paramount over the performance of control systems and of operating personnel of nuclear installation at all states considered at the nuclear installation design, and at the same moment the operating personnel shall have a possibility to put manually the protective system into operation.
§ 18
Redundancy of Protective Systems
9 (1) The protective systems shall be designed with a high functional reliability, multiplicity (redundancy) and independence of individual channels so that
a) no single failure must not result to a loss of the protective function of the system,
b) disconnection (putting out of operation) of any component or channel must not result to the reduction of number of independent (redundant) components or channels to only one, as far as there is not possible in such a case to demonstrate acceptable reliability of operation of protective system.
(2) The protective systems shall enable the periodic tests of function of individual independent channels at the reactor operation and the testing the joint circuits of the independent channels at least at the reactor shut down. These joint circuits shall be designed in such a way to assure that their possible failures may result, as a maximum, to the reactor shutdown, and in no case to the loss of the protective function.
§ 19
Relation of Protective and Control Systems
(1) The protective and control systems shall be separated thus, that any failure of control systems must not influence the capability of protective systems to perform the required safety function. Functionally necessary and purposeful interconnection of protective and control systems shall be limited to a maximum degree thus, so that it must not significantly influence the nuclear safety.
(2) The protective system shall be designed in such a way that it must not be possible to exceed the design limits even in the event of malfunction of the control system. The protective interventions are superior at all states considered in the nuclear installation design to the action of the control system and of operating personnel of the nuclear installation.
§ 20
Control Room
(1) The nuclear installation, the part of which is a nuclear reactor shall be provided with a minimum one operational control room, from which it can be safely and reliably monitored and controlled under normal and abnormal operations and at the origination of accident conditions.
10 (2) The operational control room shall be designed in such a way, that from the viewpoint of the operating personnel protection it may enable the access, safe stay and its health non-defectiveness even under the accident conditions.
(3) The nuclear installation shall enable the reactor shutdown and the reactor keeping in a safe state, even if the operational control room becomes unserviceable. The appropriate component shall be sufficiently, physically and electrically, separated from the operational control room.
§ 21
Reactor Shutdown
(1) The reactor shall be provided with the systems which are capable to shut down it under normal and abnormal operations and under the accident conditions. They shall ensure that the shut down can be maintained even at the most reactive core situation. The effectiveness, speed of action and shutdown margin shall assure that the specified design limits are not exceeded.
(2) The reactor shutdown means shall be consist of, as a minimum, two independent systems based on different principles and been capable to perform their functions even in the case of a single failure.
(3) At least one of the systems mentioned in paragraph (2) shall be self-supportingly capable to put quickly the reactor from normal or abnormal states and from accident conditions into the subcritical state with a sufficient margin assuming a simple failure.
(4) At least one of the systems mentioned in paragraph (2) shall be self-supportingly capable to put reactor from the normal operation into the subcritical state and of maintaining the reactor in a subcritical stage with a sufficient margin even at the most reactive core situation.
(5) In proving the required properties of means for the reactor shutdown, the special attention shall be devoted to the failures originating anywhere at nuclear installation that might put out of operation a part of these components.
(6) The means for the reactor shutdown shall be capable to prevent the spontaneous origination of criticality. This requirement shall be fulfilled even under activities leading to the reactivity increase during the shut down reactor (for example, withdrawal of absorber
11 element for the purpose of maintenance or fuel reload), namely even providing single failures of these shutdown means.
(7) The instrumentation systems and tests shall assure that the means for reactor shutdown are at required state.
(8) A part of these reactor shutdown means can be used during normal operation for the reactivity control or for the neutron flux shaping, as far as shutdown margin is continuously kept.
PART V
Reactor Cooling Systems
§ 22
Principles for Primary Circuit Design
(1) The primary circuit and its auxiliary, control and protection systems shall be designed in such a way
a) to assure with a sufficient margin under normal and abnormal operations the required strength, service life-time and functional reliability of their parts and components,
b) to prevent non-allowable coolant leakage,
c) to have sufficient resistance against the occurrence and development of failures and to assure the slow development of appropriate failures and their timely identification,
d) to exclude the failures of large extent,
e) to assure that the start of action of pressure relief devices (safety valves) will not cause non-allowable leakage of radionuclides of the nuclear installation,
f) to assure that components of primary circuit containing the coolant, as the pressure vessel, pressure piping, tubes and their connections, valves, sealing etc. including their fixings, will withstand static and dynamic loads anticipated during all operational states and during accident conditions.
(2) The design of components of primary circuit shall
a) determine the materials verified for these purposes and corresponding to the appropriate rules, technical standards or technical conditions,
12 b) prove by the theoretical computation and by experimental verification their sufficient sizing,
c) include the margin on the deterioration of material properties that may originate during the operation owing to a erosion, corrosion, material fatigue, chemical conditions, irradiation and ageing, and the margin on the uncertainties in the determination of the components initial states and of rate of the deterioration of their properties,
d) content the analyses of limiting states with regard to the origination and propagation of cohesion (integrity) failures,
e) stipulate the manner of evidence of the production and mounting quality by accessible up-to-date methods, and stipulate the manner of evidence of required tightness,
f) stipulate the programme and methods of determining its state during the operation.
(3) The design of primary circuit components shall content the conditions of its tests and maintenance, the conditions of normal and abnormal operations, accident conditions, the analysis and solution of all effects damaging this component.
(4) The design of primary circuit components shall content the measures for keeping the coolant amount or the coolant pressure in such a way, in order that the stipulated design limit must not be exceeded at none of states of normal and abnormal operations considering volume changes and leakage.
(5) The systems, assuring the keeping the coolant amount or the coolant pressure, shall have the adequate capacity (flow or volume) to comply with the requirements according to § 22, par. 4 and § 24, par. 1.
§ 23
In-service Primary Circuit Inspection
(1) The components of the primary circuit shall enable during the whole operational life of the nuclear installation to carry out periodically or continuously the inspection of their states during the operation (i.e. in-service inspection) and tests necessary for the nuclear safety verifications.
(2) The parts of the primary circuit design are
a) the in-service inspection programme and diagnostic procedures,
b) the criteria for the evaluation of inspection and of results of tests.
13
§ 24
Coolant Refilling and Purification System
(1) The coolant refilling system shall be capable to compensate leakage and fuel volume changes under normal and abnormal operations, considering the coolant bypass for purification, in order that the stipulated design limits may be kept.
(2) The coolant purification system shall be capable to remove the corrosion products and fission products which release from the fuel elements at their appropriate damages, and in the same time to keep the required purity parameters of the primary circuit.
§ 25
Residual Heat Removal System
(1) The residual heat removal system shall assure that during the reactor shutdown the design limits of the fuel elements and of the primary circuit must not be exceeded.
(2) The residual heat removal system shall provide the sufficient redundancy of important components of the residual heat removal system, the suitable interconnection, the capabilities for disconnection of parts of system, the leakage detection and the capability of their retaining thus, in order that the system may reliable operate even at a single failure.
§ 26
The emergency core cooling system shall assure
a) reliable cooling the core under the accident conditions caused by a loss of coolant, that
- 1. the temperatures of fuel cladding must not exceed the values stipulated by design limits,
- 2. the energy contribution of the chemical reactions (cladding, water, hydrogen release) must not exceed the acceptable value,
- 3. the changes of the fuel elements, fuel assemblies and reactor internals, which could influence the efficiency of the cooling, must not originate,
- 4. the residual heat may be removed for a sufficiently long period,
b) its sufficient redundancy, suitable interconnection, capability of disconnection of parts of system, leakage detection and the capability of their retaining thus, in order that the system may reliable operate even at a single failure.
14
§ 27
In-service Inspection of Emergency Core Cooling System
The emergency core cooling system shall enable to carry out the periodic tests and inspections of
a) resistance and tightness of system,
b) active components of system and the verification of their functionality,
c) the emergency cooling as a whole and the verification of its functionality under the conditions corresponding to its operation (sequence of operations which put the individual components into operation, switching to spare (emergency) power sources or to other cooling water system etc).
§ 28
Secondary Circuit
The secondary circuit design shall assure:
a) the reliable heat removal from the primary circuit,
b) the identification of appropriate leakage from the primary into the secondary circuit, and if those leakage are found, it shall enable the limitation of their further propagation in such a way, in order that the stipulated limits of radionuclide escapes into the environment must not be exceeded.
PART VI
Power Supply Systems
§ 29
Power Supply Systems
(1) The power output of a nuclear installation and internal load supply shall assure that
a) their external and internal failures of power supply may affect the reactor operation and the heat removal systems as little as possible,
b) the plant components important for operation may be powered from two different sources (on-site generator and electricity transmission network).
15 (2) The electric power supply to the control and protective systems of the primary circuit, the residual heat removal systems, the emergency cooling system and to the Containment System shall also enable the power supply from an emergency source, i.e. to be redundant without any power limitation for the time necessary for reliable function of systems, and independently on the reality if in operation there are on-site generators or off-site electricity transmission network. The control and protective systems shall be continuously powered.
§ 30
Redundancy of Power Supply Systems
(1) The systems, that are with regard to the nuclear safety redundant, shall be powered in a such way that their functional independence may be ensured by the way, that the electrical power systems and their sources are reciprocally independent. If the number of sources is lower than the number of independent systems, the design shall demonstrate that this does not reduce their reliability.
(2) If a single failure of powered systems cannot affect their function, a single failure of electrical system or source is also admitted.
(3) As far as the operational capability of some system is necessary for the nuclear safety assurance, the electric power supply system shall provide the necessary power even at a single failure, without limitations.
§ 31
Emergency Power Sources
(1) The systems which shall be powered without any interruption (the I st category consumers), shall be supplied from the sources that provide the power immediately (batteries with alternators).
(2) The sources and supply systems which are only put into operation after a certain duration of accident conditions (the II nd category consumers) shall be put into operation within time shorter than it is the start-up period of the II nd category consumers.
(3) The capability to carry out the functional testing the emergency power sources shall be assured.
16 PART VII
Containment System
§ 32
Purpose and Significance
The nuclear installations, the part of which is the reactor with a power above 50 MWt, shall be provided with the Containment System which, in the event of accident conditions, including the maximum design basis accident, connected with the leakage of radionuclides and ionising radiation released by them, limits such leakage into the environment in such a way, that they may be in accordance with special Regulation 1), as far as this function is not assured by other technical means.
§ 33
Design Principles
(1) The Containment System consists of the hermetic envelope dimensioned for all design basis accidents, of the closing elements, of the pressure and temperature reduction systems and of venting and filtration systems.
(2) The Containment System shall assure in order that its required tightness may be kept under the origination of accident conditions and a sufficiently long time after their termination.
(3) The Containment System shall render the required function for maximum pressures and appropriate under-pressures and temperatures of design basis accidents. There is necessary to consider the influence of the pressure and temperature reduction systems within a the protective envelope, the influence of other potential power sources, penetrations and access openings, the uncertainties of the calculation models, the experimental results and the operating experience.
(4) The Containment System shall fulfil the requirements for the protection against external effects according to § 10.
(5) The components of the Containment System shall assure their functional capability and limit the influence on other systems and components important from the viewpoint of nuclear safety.
§ 34
17 Inspection of Hermetic Envelope Tightness
(1) The hermetic envelope and components important for its tightness shall assure that it may be possible
a) to carry out the leakage rate test under the design pressure with all penetrations and hatches already installed,
b) to perform the in-service periodic inspections of the tightness of the Containment Systemunder the design pressure or under lower pressures which will allow the extrapolation.
(2) The hermetic envelope and components important for its tightness shall enable the leakage rate tests of performed repairs under the design pressure.
§ 35
Hermetic Envelope Pressure Test
The hermetic envelope shall enable before the nuclear installation commissioning to demonstrate by the leakage rate test its integrity under the test pressure which is higher than the design one.
§ 36
In-service Inspection of Containment System
The Containment System shall enable so that during the nuclear installation operation they may be carried out
a) the periodic inspections of its individual parts and components,
b) the functional tests of its individual parts and components.
§ 37
Hermetic Envelope Walls Penetrations
The pipelines and cables penetrations passing through the walls of hermetic area are designed in such a way, so that
a) there may be carried out the leakage detection, their capturing and collection,
b) there may be carried out the regular tests of their tightness under the design pressure independently on the hermetic envelope leakage rate tests,
18 c) there may be assured their protection against effects of dynamic forces.
§ 38
Closing Elements
(1) Primary circuit pipelines which pass through the walls of hermetic envelope, or pipelines which are directly connected with the atmosphere within the hermetic envelope shall be provided with reliable closings, each of them shall have at least two closing elements being arranged in series that shall be located outside and inside the hermetic envelope and are independent and reliably controlled. The external closing elements are located as close to the hermetic envelope as practical.
(2) Other pipelines penetrating the walls of hermetic envelope shall be provided with at least one external closing element which shall be located as close to the hermetic envelope as practical.
(3) The closing elements shall be designed in such a way so that the leakage rate tests may regularly carry out.
(4) The functional capability of the closing element shall be assured even assuming a single failure out of its mechanical part.
§ 39
Hermetic Area Access Openings
The operational access openings through the hermetic envelope shall be equipped as the closings by double doors controlled in such a manner, that their tightness may be always assured. The tightness of assembling access openings shall correspond to the tightness of the Containment System.
§ 40
Internal Arrangement of Hermetic Area
Among the individual parts of hermetic area there shall be sufficient ducts, so that the pressure gradients originated under the accident conditions must not damage the hermetic envelope or other components of the hermetic envelope system.
§ 41
Hermetic Envelope Pressure Reduction and Heat Removal System
19 (1) The hermetic area shall be provided with the pressure reduction and heat removal system which, together with other systems, may after the termination of accident conditions, connected with releases of mass and energy, sufficiently assure the fast reduction of pressure and temperature in the hermetic area, and which further on may assure that their admissible values must not be exceeded.
(2) The system shall assure the reliability, redundancy and functional heterogeneity of its important components and assure its functional capability, even at a single failure.
§ 42
Other Systems of Protective Envelope
(1) The Containment System shall be provided with systems that assure the monitoring the fission products and substances which might enter into it at origination of accident conditions. These systems shall be capable together with other systems
a) to reduce the volume activity and to adjust fission products composition,
b) to monitor the volume concentrations of explosive substances, in order that the integrity of hermetic envelope may be assured and in order that the amounts of releasing radionuclides may be reduced.
(2) The important components of these systems shall be redundant in order that they may operate at a single failure.
PART VIII
Radiation Protection
§ 43
Ionising Radiation and Radionuclides Monitoring
The ionising radiation and radionuclides monitoring shall be assured in accordance with a special Regulation 1).
§ 44
Venting and Filtering Systems
The nuclear installation shall be equipped by
a) the venting and filtering systems which under normal and abnormal operations shall
20
- 1. prevent the scattering and non-controlled leakage of radionuclides in individual zones of nuclear installation in compliance with requirements of their accessibility,
- 2. reduce the volume activities of radionuclides under the values stipulated by the special Regulation 1) in the case of scattering and leakage of radionuclides into the zones of nuclear installation that are accessible,
- 3. keep the prescribed climatic conditions,
- 4. keep the leakage of radionuclides into the environment below the stipulated limits.
b) the reliable filters with the sufficient retention efficiency and enable the tests of their capability,
c) the redundancy of important components, so that the venting systems may operate at a single failure.
§ 45
Radionuclides Effluents into Environment
The radionuclides effluents into the environment shall fulfil the requirements stipulated by a special Regulation 1).
PART IX
Nuclear Fuel Handling and Its Storage
§ 46
Fresh Nuclear Fuel Handling and Its Storage
The installation for handling with the fresh nuclear fuel and for its storage shall
a) prevent with margin the achievement of criticality, even under conditions of the most effective moderation of neutrons (optimum moderation) by the geometrically safe configuration or by other physical means and procedures, and by this to prevent
- 1. the exceeding the 0.95 value of effective neutron multiplication coefficient under the assumed accident situations (including the flooding by pure water),
- 2. the exceeding the 0.98 value of effective neutron multiplication coefficient under the conditions of optimum moderation,
21 b) assure the capability for performance of periodic inspections and tests,
c) reduce to the minimum the possibility of damage or loss of fuel,
d) prevent fall of fuel during the transport,
e) prevent the fall of heavy objects on the fuel assembly, i.e. objects with a mass greater than the mass of fuel assembly,
f) assure the storage of elements or fuel assemblies at constructions and operational units the part of which is a nuclear reactor.
§ 47
Irradiated and Spent Nuclear Fuel Handling and Its Storage
The installation for the handling with the irradiated and spent nuclear fuel and its storage, and for the handling and storing the other substances containing the fissile products and radioactive substances shall be designed in a such way, in order that it may be possible
a) to prevent with margin the achievement of criticality even under conditions of the most effective deceleration of neutrons (optimum moderation) by area arrangement or by other physical means and procedures, and by this to prevent
- 1. the exceeding the 0.95 value of effective neutron multiplication coefficient under the assumed accident situations (including the flooding by water),
- 2. the exceeding the 0.98 value of effective neutron multiplication coefficient under the conditions of optimum moderation,
b) to assure the adequate residual heat removal under normal and abnormal operations and under accident conditions,
c) to assure the capability for performance of periodic inspections and tests,
d) to prevent the fall of irradiated fuel during the transport,
e) to reduce to the minimum the possibility of fuel damage, i.e. namely to prevent the exposure of irradiated element or fuel assembly to the non-allowable load during the handling,
f) to prevent the fall of heavy objects on the fuel assembly, i.e. the objects with the mass greater than the mass of fuel assembly is,
22 g) to enable the storage of damaged fuel elements or damaged fuel assemblies at the constructions and operational units, the part of which is a nuclear reactor,
h) to assure the radiation protection of nuclear installation personnel,
i) for wet storage with a water charge to assure
- 1. the check-up of chemical composition and of radioactivity of all water, inside of which the irradiated fuel is stored or in which there is a handling with it,
- 2. the monitoring and controlling the height of water level in the spent fuel pool and the leakage detection.
PART X
Final Provision
§ 48
Force
This Regulation comes into force by the day of proclamation.
In g. K areO% KP
Acting Chairman of State Office for Nuclear Safety
- 1) The Regulation No 184/ 1997 Sb. of the State Office for Nuclear Safety, on Requirements on Assurance of Radiation Protection.
- 2) Regulation No. 219/1997 Sb. of the State Office for Nuclear Safety, on Particulars for Assurance of Emergency Preparedness of Nuclear Installations and Workplaces with Ionising Radiation Sources and on Requirements for Content of On-site Emergency Plan and the Emergency Code.
- 3) Governmental Decree No. 11/1999 Sb., on the Emergency Preparedness Zone.
- 4) For example, the Law No. 133/1985 Sb., on Fire Protection, in last wording.
23
- 5) Regulation No. 144 / 1997 Sb. of the State Office for Nuclear Safety, on the Physical Protection of Nuclear Material and Nuclear Installations and on Its Ranking into Particular Categories.
24