German codes and standards for nuclear power plant components, recent KTA activities for advanced reactor types

Int. J. Pres. Ves. & Piping 37 (1989) 113-126

German Codes and Standards for Nuclear Power Plant Components, Recent KTA Activities for Advanced Reactor Types

K. Bieniussa, J. F r e u n d & H. R e c k Geseilschaft fiir Reaktorsicherheit(GRS) mbH, Schwertnergasse 1, D-5000 K61n 1, FRG

ABSTRACT The German codes and standards for the design of nuclear power plant components are the KTA Safety Standards. For comparison with international code activities, thefollowing aspects are presented: organization and function of the German Nuclear Safety Standards Commission ( KTA ), present extent of standard activities, standards for metallic components, bases for the development of standards, comparison with international standards, and current standardization activities for advanced reactor types. As far as the Fast Breeder Reactor (FBR) is concerned, the designers of metallic components deal with the refinement of their SNR-3OO-specific design criteria. With respect to the High Temperature Reactor ( HTR ), all parties interested in the improvement of the HTR concepts cooperate in the preparation of KTA Safety Standards for metallic, ceramic and concrete components.

1 INTRODUCTION The German codes and standards for nuclear power plants are the KTA Safety Standards. They are prepared by special working groups which are established by the German Nuclear Safety Standard Commission (KTA). The standards for Light Water Reactors (LWRs) are nearly complete. In 1979 a group consisting of various institutions began the development of safety criteria for High Temperature Reactor (HTR) components, considering material behaviour and design aspects. This project was 113 Int. J. Pres. Ves. & Piping 0308-0161/89/$03-50 © 1989 Elsevier Science Publishers Ltd, England. Printed in Great Britain

114

K. Bieniussa, J. Freund, H. Reck

sponsored by the government of the Federal Republic of Germany (FRG). Now the KTA is assessing the proposed safety criteria for final formulation of adequate HTR safety standards. This paper deals with the following aspects: --Licensing procedure in the FRG, --Organization and function of the KTA, --Standardization activities for advanced reactor types, - - K T A activities for HTR safety standards.

2 LICENSING P R O C E D U R E IN THE FEDERAL REPUBLIC OF GERMANY In the FRG the generation and peaceful utilization of nuclear power is considered in the Constitutional Law. The Atomic Energy Act 1 describes basic requirements for the design, construction and operation of nuclear installations (Fig. 1). The Radiation Protection Ordinance 2 defines the maximum admissible release of radioactive products. With respect to components the principles of the Safety Criteria 3 demand precautions against failure, for example by: ---consideration of adequate safety factors for the component design; --material qualification as well as adequate quality assurance in connection with fabrication, construction and operation; --simple and reliable components with good repairability; --inservice inspection for early detection of defects. Further licensing preconditions are required, e.g. in RSK Guidelines ( R S K = Reactor Safety Commission) 4 for Pressurized Water Reactors (PWRs) and KTA Safety Standards which are valid for all LWRs. Corresponding standards for HTR plants are in preparation, or may be derived from these principles.

Safety Criteriak

fetyStandardsand~

Issuedby Authoritmsk Fig. 1. Classification scheme for regulations and standards.

Rulesand Guides pecifications

k

German codes and standards for nuclear plant components

115

The licensing procedure for construction and operation of a Nuclear Power Plant (NPP) will be carried out by the authority of the state in which a NPP will be installed. The state authorities are supervised by a federal authority giving uniform interpretation and application of safety regulations throughout the country. To obtain a licence it must be shown in the safety report that the above mentioned precautions are fulfilled and that the possible release of radioactive fission products is below specified limits. Additional documents are submitted to the state authority, e.g. documents in which safety related problems and their solutions are dealt with in detail. After approval of these necessary documents the authority will give a first partial licence for the start of construction work. The succeeding partial licences will be given stepwise, corresponding to the particular construction procedure. After final completion and successful performance of functional tests the licence for continuous operation is the last which will be given. This licensing procedure is common in the FRG for LWRs and has been carried out successfully also for the HTRs AVR (Arbeitsgemeinschaft V ersuchsreaktor), THTR (T_horium Hochtemperatur-_Reaktor) and the Fast Breeder Reactor (FBR) KNK. Construction of the FBR SNR-300 is finished and now the plant is waiting for the operational licence.

3 O R G A N I Z A T I O N AND FUNCTION OF THE KTA The Nuclear Safety Standards Commission (KTA) was instituted 1972 by the Government of the Federal Republic of Germany. 5 It is made up of five groups with ten appointed members each, representing the nuclear parties concerned: the manufacturers, the utilities, the licensing and supervising authorities, the safety reviewing organizations, and finally a group which represents various parties taking an interest in the nuclear field. The mandate of the KTA is to develop safety standards and promote their use in those areas of nuclear technology where, on the basis of experience, a consensus is expected from experts representing manufacturers, utilities, regulatory authorities, and safety reviewing organizations. 6 The KTA's responsibility for the entire field of nuclear safety technology quite obviously does not mean it has to simultaneously set up safety standards for the entire field. According to the advantages that can be achieved through the establishment of standards in view of the number of facilities to be licensed, currently mainly safety standards for LWR nuclear power plants have been established. Up to the time of writing 68 standards have been published covering the following topics: administration and organization, quality assurance, radiation protection and monitoring, internal events, external events, structural engineering, reactor core and

Requirements for the Operating Manual (replaces version of Mar. 1981) Requirements for the Testing Manual Radiological Protection Considerations for Plant Personnel in the Design and Operation of Nuclear Power Plants; Part l: Design Radiological Protection Considerations for Plant Personnel in the Design and Operation of Nuclear Power Plants; Part 2: Operation General Requirements for Quality Assurance (replaces version of Feb. 1980) Quality Assurance for Weld Materials and Weld Additives for Pressure and Activity Retaining Components in Nuclear Power Plants; Part 1: Suitability Testing Quality Assurance for Weld Materials and Weld Additives for Pressure and Activity Retaining Components in Nuclear Power Plants; Part 2: Manufacture Quality Assurance for Weld Materials and Weld Additives for Pressure and Activity Retaining Components in Nuclear Power Plants; Part 3: Processing Stationary System for Monitoring Local Dose Rates within Nuclear Power Plants Monitoring Radioactivity in the Inner Atmosphere of Nuclear Power Plants; Part 1: Nuclear Power Plants with Light Water Reactors Measuring and Monitoring the Discharge of Gaseous and Aerosolbound Radioactive Materials; Part 1: Measuring and Monitoring the Stack Discharge of Radioactive Materials during Specified Normal Operation

KTA 1 201 KTA 1 202 KTA 1 301.1

KTA 1 503.1

KTA 1 502.1

KTA 1 501

KTA 1 408.3

KTA 1 408.2

KTA 1 408.1

KTA 1 401

KTA 1 301.2

Title

Safety standard

TABLE 1 K T A Standards and Draft Standards

R

R

R

R

(Feb. 1979)

(June 1986)

(Oct. 1977)

(June 1985)

(June 1985)

(June 1985)

R

R

(Dec. 1987)

(June 1982)

R

R

R (Nov. 1984)

R (Dec. 1985) R (June 1984)

issue a

Status and date of

av

av

ip

ip

ip

ip

av

av

av av

Translation b

B

C

B

B

B

B

A

A

B

B A

Applicable for H T R c

e~

e~ 9-

~"

KTA 3 102.3

KTA 3 102.2

KTA 3 102.1

KTA 3 101.2

KTA 2 207 KTA 2 501 KTA 3 101.1

KTA 2 201.5

KTA 2 201.4

KTA 2 201.2

KTA 2201.1

KTA 2 101.1 KTA 2 103

KTA 1 508

KTA 1 507

KTA 1 506

KTA 1 504

Measuring Liquid Radioactive Materials for Monitoring the Radioactive Discharge Measuring Local Dose Rates in Exclusion Areas of Nuclear Power Plants Monitoring the Discharge of Gaseous, Aerosol-bound and Liquid Radioactive Materials from Research Reactors Instrumentation for the Determination of the Dispersion of Radioactive Materials in the Atmosphere Fire Protection in Nuclear Power Plants; Part 1: Basic Principles Explosion Protection in Nuclear Power Plants with Light Water and High-Temperature Reactors Design of Nuclear Power Plants against Seismic Events; Part 1: Basic Principles Design of Nuclear Power Plants against Seismic Events; Part 2: Soil Foundation Design of Nuclear Power Plants against Seismic Events; Part 4: Design of Mechanical and Electrical Components Design of Nuclear Power Plants against Seismic Events; Part 5: Seismic Instrumentation Protection of Nuclear Power Plants against Floods Sealing of Structures in Nuclear Power Plants Reactor Core Design for Pressurized and Boiling Water Reactors; Part 1: Thermohydraulic Design Reactor Core Design for Pressurized and Boiling Water Reactors; Part 2: Requirements with Respect to Neutron Physics for the Design and Operation of the Reactor Core and Neighbouring Systems Reactor Core Design for High-Temperature Gas-Cooled-Reactors; Part 1: Calculation of the Material Properties of Helium Reactor Core Design for High-Temperature Gas-Cooled-Reactors; Part 2: Heat Transport in Spherical Fuel Elements Reactor Core Design for High-Temperature Gas-Cooled-Reactors; Part 3: Loss of Pressure through Friction in Pebble Bed Cores (Sept. 1988) (Dec. 1985)

(Mar. 1984)

(June 1986)

(June 1978)

(Nov. 1982)

(June 1975)

R

R

R

(Mar. 1981)

(June 1983)

(June 1978)

(Dec. 1987)

(Feb. 1980)

R

R

(June 1977) (June 1982) (Sept. 1988)

R R R

RE (Nov. 1983)

R

R

RE (June 1985~

R R

R

R

R

ip

ip

av

ip

av av

ip

av

ip

ip

av

~, ominued~

A

A

A

C

C

A A

R

A

A

B

B

"....1

E"

e~

",t

%

Safety

KTA 3 401.2

KTA 3 301 KTA 3401.1

KTA 3 204 KTA 3 205.1

KTA 3 203

KTA 3 201.4

KTA 3 201.3

KTA 3 201.2

KTA 3 103 KTA 3 104 KTA 3 201.1

KTA 3 102.5

KTA 3 102.4

standard

Reactor Core Design for High-Temperature Gas-Cooled-Reactors; Part 4: Thermohydraulic Analytical Model for Stationary and QuasiStationary Conditions in Pebble Bed Cores Reactor Core Design for High-Temperature Gas-Cooled-Reactors; Part 5: Systematic and Statistical Errors in the Thermohydraulic Core Design of the Pebble-Bed Reactor Shutdown Systems of Light Water Reactors Determination of the Shutdown Reactivity Components of the Reactor Coolant Pressure Boundary of Light Water Reactors; Part 1: Materials (replaces version of Feb. 1979) Components of the Reactor Coolant Pressure Boundary of Light Water Reactors; Part 2: Design and Analysis (replaces version Oct. 1980) Components of the Reactor Coolant Pressure Boundary of Light Water Reactors; Part 3: Manufacture (replaces version of Dec. 1979) Components of the Reactor Coolant Pressure Boundary of Light Water Reactors; Part 4: Inservice Inspections and Operational Monitoring Monitoring Radiation Embrittlement of Materials of the Reactor Pressure Vessel of Light Water Reactors Reactor Pressure Vessel Internals Component-Support Structures with Non-Integral Connections; Part 1: Component-Support Structures with Non-Integral Connections for Components of the Reactor-Coolant Pressure Boundary Residual-Heat Removal Systems for Light Water Reactors Steel Containment Vessels; Part 1: Materials (replaces version of Nov. 1982) Steel Containment Vessels; Part 2: Analysis and Design (replaces version of June 1980)

Title

T A B L E l--contd.

av

ip

R (Sept. 1988) R (June 1985)

ip ip

R (June 1982) R (Nov. 1984)

C

C

C

B

B

C

ip

B

B

B

B

A C C

A

Applicable for H T R c

av

ip

ip ip

av

ip

Translation b

R (Mar. 1984) R (Mar. 1984)

R (June 1982)

R (Dec. 1987)

R (Mar. 1984)

R (Nov. 1982)

R (June 1986) R (Mar. 1984) R (Oct. 1979)

R (Nov. 1984)

Status and date of issue"

e0

oo

3 502 3 503 3 504 3 505

KTA 3 507 KTA 3 601 KTA 3 602

KTA 3 506

KTA KTA KTA KTA

KTA 3 501

KTA 3413

KTA 3 407 KTA 3 409

KTA 3 405

K T A 3404

KTA 3 403

KTA 3 40 1.4 KTA 3 402

KTA 3 401.3

Steel Containment Vessels; Part 3: Manufacture (replaces version of Dec. 1979) Steel Containment Vessels; Part 4: Inservice Inspections Air Locks through the Containment Vessel of Nuclear Power Plants-Personnel Locks Cable Penetrations through the Reactor Containment Vessel (replaces version of Nov. 1976) Closure of Pipe Penetrations through the Reactor Containment in Case of a Release of Radioactive Materials inside the Reactor Containment Integral Leakage Rate Testing of the Containment Vessel with the Absolute Pressure Method Pipe Penetrations through the Reactor Containment Vessel Air Locks through the Containment Vessel for Nuclear Power Plants-Material Locks Determination of the Loads for the Design of the Large Dry Containment against Plant-Internal Incidents Reactor Protection System and Monitoring Equipment of the Safety System (replaces version of Mar. 1977) Incident Instrumentation Type Testing of Electrical Modules of the Reactor Protection System Electrical Drives of the Safety System in Nuclear Power Plants Type Testing of Measuring Transmitters and Transducers of the Reactor Protection System System Testing of the Instrumentation and Control Equipment of the Safety System in Nuclear Power Plants Factory Tests of Equipment for Instrumentation and Control Ventilation and Air Filtration Systems in Nuclear Power Plants Storage and Handling of Nuclear Fuel Assemblies, Control Rods, and Neutron Sources in Nuclear Power Plants with Light Water Reactors (replaces version of June 1982) (June (Nov. (June (Sept.

1985) 1982) 1982) 1988)

R

(June 1984)

R (Nov. 1984) R (Nov. 1986) RE (Oct. 1979)

R (Nov. 1984)

R R R R

RE (Dec. 1985)

R (June 1979)

R (Feb. 1979) RE (June 1985)

R (Sept. 1988)

(Oct. 1980)

(Nov. 1976)

R R

(Nov. 1986) (Mar. 1981)

R R

ip

av

ip

av

ip

av

av

av

av

av

ip

(c ominued~

C

A A B

B

B B B B

C

C

C C

C

C

C

C C

Safety

Facilities for Treating Radioactively Contaminated Water in Nuclear Power Plants Storage, Handling and Plant-Internal Transportation of Radioactive Materials (Other than Nuclear Fuel Assemblies) in Nuclear Power Plants (replaces version of Nov. 1982) Basic Requirements for the Electrical Power Supply of the Safety System in Nuclear Power Plants; Part 1: Single-Unit Plants Basic Requirements for the Electrical Power Supply of the Safety System in Nuclear Power Plants; Part 2: Multi-Unit Nuclear Power Plants Emergency Power Facilities with Diesel Generators; Part 1: Design Emergency Power Facilities with Diesel Generators; Part 2: Tests and Examinations Emergency Power Facilities with Batteries and Rectifiers Emergency Power Facilities with DC/AC Converters Switching Facilities, Transformers and Distribution Networks for the Electrical Power Supply of the Safety System in Nuclear Power Plants Communication Devices for Nuclear Power Plants (replaces version of Mar. 1977) Lifting Equipment in Nuclear Facilities (replaces version of June 1978) Testing and Operation of Lifting Equipment in Nuclear Facilities Control Room, Emergency Control Room and Local Control Stations in Nuclear Power Plants

Title

(Sept. 1988) R

(Sept. 1988)

R

(Mar. 1981) (Nov. 1983) (Nov. 1982)

(Nov. 1982) (June 1986) (June 1984)

R R R

R R R

(June 1982) (June 1980)

(June 1978)

(June 1983)

(Feb. 1980)

R R

R

R

R

Status and date of issue"

ip

av

ip ip ip

av av

av

ip

Translation h

B

A B B

B B B

B B

B

A

A

Applicable for HTR"

a R E = draft Safety Standard ('Regelentwurf'); R = Safety Standard ( ' R e g e l ' ) . b ip = translation in preparation; av = translation or draft translation availablc from KTA Secretariat (c/'o GRS mbH, Schwcrtnergasse 1, D-5 000 Krln, FRG). c A applicable for HTRs without any restrictions: B = applicable for HTRs in most rcspects, only minor adaptations arc necessary to covcr HTRs in all respects: C = restricted to LWRs.

K T A 3 902 KTA 3903 KTA 3904

K T A 3 901

KTA 3703 KTA 3704 KTA 3705

KTA 3 702.1 KTA 3 702.2

KTA 3 701.2

KTA 3 701.1

K T A 3 604

KTA 3 603

standard

TABLE 1--eontd

e~

e~

%

German codes and standards for nuclear plant components

121

reactor control, primary and secondary circuits, heat removal, containment vessels, instrumentation and reactor protection, activity monitoring and ventilation, energy and media supply, and auxiliary systems. A list of standards, some available in English, is given in Table 1.

4 STANDARDIZATION ACTIVITIES FOR ADVANCED REACTOR TYPES

4.1 Fast breeder reactor design criteria The construction of the sodium cooled fast breeder reactor SNR-300 is finished, but the commissioning licence has not yet been issued (late 1988). This plant has been constructed as a loop type where the main components of the primary system are connected by pipes. The components which have to cope with temperatures up to 550°C, i.e. which are used in the creep range, have been designed according to SNR-specific design criteria. So far, these activities, which are at present only undertaken by the designers of the SNR-300 components, have not been integrated into the national standardization work pursued by KTA. Other KTA safety standards, e.g. KTA 1401 'General Requirements for Quality Assurance', are also applied to the SNR-300.

4.2 High temperature reactor design and safety criteria The helium cooled high temperature reactor (HTR) THTR-300, which achieves gas outlet temperatures of T = 750°C, was successfully commissioned in 1986. Since further advanced HTR projects of various sizes and for various purposes (with gas outlet temperatures up to T = 750°C for the generation of power, and up to T = 1000°C for the generation of nuclear process heat) have long been under discussion, there was a general interest in the preparation of HTR-specific design criteria for metallic components, the prestressed concrete reactor pressure vessel, and ceramic internal components of the reactor pressure vessel. Therefore the Nuclear Research Center at Jiilich (KFA) took the lead, when the preparation commenced of the bases for design standards of HTR components which should be used above application temperatures of 800°C. The first four-year phase, which was sponsored by the Federal Ministry of the Interior (BMI), was finished in December 1983.7 At first, all the design criteria for HTR components for applications above 800°C were developed. In a second four-year phase immediately following the first, design criteria for all HTR applications, irrespective of the type concerned (steel or prestressed concrete pressure

22

K. Bieniussa, J. Freund, H. Reck

vessel), have been and still are being developed. This project is funded by the Federal Ministry for Research and Technology (BMFT). Apart from KFA, the following organizations, manufacturers, experts and scientific institutions participate in the project: Brown Boveri & Cie, Mannheim; Gesellschaft fiir Reaktorsicherheit, K61n; Hochtemperatur-Reaktorbau, Mannheim; Interatom, Bensberg; Institut fiir zerst6rungsfreie Priifung, Saarbriicken; Materialpriifungsanstalt der Universit/it Stuttgart, Stuttgart; Rheinisch-Westf'alischer Technischer Oberwachungs-Verein, Essen; SIGRI Meitingen and Zerna, Schnellenbach und Partner, Bochum. The development of safety and design criteria for HTR components involves four working domains (Fig. 2). In the first step only the components important for safety are considered. In further steps the other safety related components are considered. In domain A the overall safety concept is formulated by defining the necessary safety principles, the integrity requirements for the components, and the assumptions for design related incidents and accidents. In domain B all the metallic components in contact with the primary cooling gas are considered. Here the design criteria with regard to materials, design, manufacturing, and inspection are elaborated. Domain C deals with prestressed concrete reactor pressure vessels including the gas tightening steel liner, and thermal insulation and liner cooling system, and the concrete or steel closures of the vessel. Criteria are formulated for design and manufacture aspects. In domain D the graphite reactor components are covered with respect to radiation and corrosion influences, as well as design and construction aspects. Graphite reactor components must be designed having consideration for failure probability. This is the principal difference from the design of metallic components which uses safety factors to avoid component failure, s

DOMAIN A Safety Criteria

DOMAIN B Metallic Components

DOMAIN C I Prestressed Con- [ crete Reactor Pressure Vesse

I Safety Principles,

Materials, Manufacturing, Constitutive Equations, Construction,

Integrity,

Loading Categories,

Assumptions for Accident-Analyses

Design by Analyses, Inspection

Fig. 2.

Prestressed

Concrete Structures,

j

DOMAIN D Graphite Reactor Components

--7

Materials, Corrosion, Loading Categories,

Liner, VesselClosings,

Construction, Design by Analyses

Thermal Isolation

Programme of development for HTR design criteria.

German codes and standards for nuclear plant components [ - -

I

DOMAIN B Metallic Components

DfOMAINA ety Criteria

I

i

DOMAIN C

123

F-

I

I

Prestressed Con [ crete Reactor Pressure Vesse s

DOMAIND Graphite Reactor Components

KTA 3221 KTA 3201 TRD, AD ASME CC N47 RCC - MR Design with \ , Design with \ ~ ( Design with ' ty Factors ~ ,Safety Factors }{ Consideration of r ,,Pailure Probability

Fig. 3. Preparation scheme of KTA safety standards for HTRs. This R & D work has been finished in 1988. Up to the time of writing the KTA has been assessing the proposed safety criteria, and new KTA working groups are engaged in formulating safety standards for HTRs. The relations between these four domains, the national and international standards and the intended new KTA Standards are given in Fig. 3. Examples of this work are shown in Figs 4 and 5. 9 Domain B deals with the metallic components, including design by analysis. Here the criteria (Fig. 4) for a temperature and time dependent stress intensity limit (St) are elaborated. These criteria are being discussed and are proposed for a future safety standard. In domain D the graphite reactor components are covered. Figure 5 shows the dependence of failure probability on safety factor for graphite. It shows that for a given safety factor the failure probability has a variance of several orders of magnitude. The assumption of high safety factors for normal graphite, though in accordance with the design by analysis of metallic components, does not take into consideration the high quality of

safety of:

criteria:

stress:

S u = Rm,t,T/1.35

time:

St2 = Rm,t/3.00,T

m a t e r i a l behavior:

St3 = Rm,t,T+15K

strain:

S~=Rp,l.0,t, T

S t = MIN. I Stt, St2, St3, Stg

Fig. 4.

Criteria for a stress intensity limit S t dependent on temperature and time--proposal.

124

K. Bieniussa, J. Freund, H. Reck

5 " P ~

I 10.4 n o r m a l graphite D., 10.~

distribution

Itreultb parameter

5

reactor graphite

,..\

) J

iO-If-5

10-4 t

45

2

2,5

safety factor

Fig. 5.

Dependence of failure probability on safety factors for graphite--proposal.

reactor graphite. The use of the method of safety factors therefore is not suitable for reactor graphite materials. The method used should be one that is applied to ceramic materials. With this method an allowable stress is determined that causes a failure probability below a prescribed value. This value must be set in a safety standard. The failure probability in the section of interest can be represented by the Weibull distribution. This method of analysis is acceptable and has been confirmed in several experiments on specimens of suitable shapes. This method is proposed for a future safety standard.

5 KTA ACTIVITIES IN ESTABLISHING KTA STANDARDS FOR HIGH T E M P E R A T U R E REACTORS In accordance with experience gained with construction and operation of AVR and THTR-300 and with the development of design criteria the Nuclear Safety Standards Commission instituted a new HTR subcommittee (HTR-SC) in November 1984. This HTR-SC had been instructed to initiate and coordinate the process of standardization in the field of HTRs as well as reviewing first drafts of HTR standards before submission to the KTA. Up to the time of writing the KTA has established 68 standards and 11

German codes and standards for nuclear plant components

125

draft standards, and five of these standards were specially established for HTRs. Of the remaining 63 standards 35 standards (and five draft standards) are likewise applicable to LWRs and HTRs. Despite the fact that so many standards and draft standards are applicable to HTRs the HTR-SC arrived at the decision to review all these standards with regard to HTRs. This seemed to be reasonable due to the facts that with the standardization work the emphasis was put on LWRs, and that the working groups came from the light water reactor field. The result of the review was: --16 standards and one draft standard are applicable to HTRs without any restriction; --28 standards and five draft standards are applicable to most aspects-only with minor items should some improvements be made for optimum adaptation to the requirements of HTRs; - - t h e rest of the standards are restricted to LWRs--their subjects cover the primary loop and the containment. Forty-one standards are still in preparation. These standards will be reviewed by the HTR-SC after their completion. The HTR-SC felt it necessary to initiate work on special H T R subjects. Therefore, by the recommendation of the HTR-SC, the KTA decided to let preliminary reports be prepared by working groups for four new subjects. Preliminary reports will cover a collation of existing material (including experience from licensing and operation), the assessment of the material, and the scope of future standards. Preliminary reports will be made for: KTA 3106 KTA 3221 KTA 3231 KTA 3232

Nuclear design of reactor core for high temperature gas cooled reactors and determination of shutdown reactivity Metallic components of high temperature gas cooled reactors Safety Requirements for the Design of Prestressed Pressure Vessels of High Temperature Gas Cooled Reactors Ceramic internal components of high temperature gas cooled reactors

REFERENCES 1. Bundesminister des Innern (BMI). Bekanntmachung der Neufassung des Gesetzes fiber die friedliche Verwendung der Kernenergie und den Schutz gegen ihre Gefahren (Atomgesetz), BMI, Bonn,_!985. 2. Bundesminister des Innern (BMI). Verordnung fiber den Schutz vor Sch~iden durch ionisierende Strahlen (Strahlenschutzverordnung), BMI, Bonn, 1976.

126

K. Bieniussa, J. Freund, H. Reck

3. (a) Bundesminister des Innern (BMI). Sicherheitskriterien ffir Kernkraftwerke, BMI, Bonn, 1977. (b) Bundesminister des lnnern (BMI). Interpretationen zu den Sicherheitskriterien ffir Kernkraftwerke, BMI, Bonn, 1979. (c) Bundesminister des Innern (BMI). Sicherheitskriterien ffir Kernkraftwerke mit gasgekiihltem Hochtemperaturreaktor, Entwurf, September 1980. 4. Gesellschaft fiir Reaktorsicherheit (GRS). RSK-Leitlinien fiir Druckwasserreaktoren sowie Rahmenspezifikation Basissicherheit, GRS, K61n, 1981. 5. Bekanntmachung fiber die Bildung eines Kerntechnischen Ausschusses vom 1. Sept. 1972 (Banz. Nr. 172 vom 13.09.1972); und Bekanntmachung fiber die Obernahme des Kerntechnischen Ausschusses in die Zustfindigkeit des Bundesministers ffir Umwelt, Naturschutz und Reaktorsicherheit vom 01. Sept. 1986 (Banz. Nr. 183 vom 02.10.1986). 6. Freund, J., Philip, G. & Schwarzer, W., The Nuclear Safety Standards Commission of the Federal Republic of Germany. Nuclear Safety, 25 (1985) 623 32. 7. KFA Jfilich. Erarbeitung yon Grundlagen zu einem Regelwerk fiber die Auslegung von HTR-Komponenten fiir Anwendungstemperaturen oberhalb 800°C, March 1984; Sonderforschungsvorhaben SR 191 des Bundesministers des Innern, Final Report November 1979-December 1983. 8. KFA-Jiilich. Auslegungskriterien fiir hochtemperaturbelastete metallische und keramische Komponenten sowie des Spannbetonreaktordruckbehfilters zukfinftiger HTR-Anlagen--gefSrdert durch den Bundesminister ffir Forschung und Technologie, Statusbericht Januar 1984 bis Dezember 1985. 9. Endbericht zum Verbund-Forschungsvorhaben des BMFT: Auslegungskriterien ffir hochtemperaturbelastete metallische und keramische Komponenten sowie des Spannbetondruckbeh~ilters zukfinftiger HTR-Anlagen. Band I, IIa, lib, III, IV, Kernforschungsanlage Jfilich GmbH (lnstitut ffir Reaktorwerkstoffe) August 1988.