Summary

The main objective of the project was to implement recommendations of type II given in the Council Report on Nuclear Safety on the context of Enlargement. To accomplish this main objective the following activities were performed:

• Stress state reassessment based on less conservative material data from experiments on materials irradiated in service as part of the Periodic Safety Review Report.
• Modification of inspection procedures and their frequency based on reassessment results.

The VVER 440 Model 230 and 213 reactors represent the first and second generation of PWR designed in OKB Gidropress Podolsk. The design lifetime of the Model 230 is nearing completion in Russia/>, Finland/> and Bulgaria/> and the age of Model 213 is approximately 20 years of operation in Russia/>, Ukraine/>, Hungary/> and
Czech/> Republic/>/>.

The complexity of the problem is explained by the fact that the reactor internals of Model 230 were designed without the necessary regulatory basis while for the second generation the related standard has been issued in 1972. Unfortunately the impact of the basic operating effects (neutron irradiation, water environment, residual stresses, thermal ageing) was not properly taken into account.

The strength analyses of reactor internals for PLEX purposes in general have analogy to current standards as ASME Code Section III or PNAE G-7-002-86. Only the properties of the material caused by operating factors must be considered in terms of their impact on the procedures of respective strength analyses (static, cyclic, brittle fracture) and the probability of damage during operation.

Results of the preliminary re-assessment of the RPV internals stress state based on the first values of real service irradiated mechanical properties after 15 years of operation may be summarised as follows:

For the re-assessment the following loading cases shall be taken into account,

· normal operation conditions including safe shutdown earthquake,

· thermal stresses,

· double ended guillotine break,

· flow-structure interaction,

The related methodology for the re-assessment has been developed. The needed input quantities are:

· the curve K1 versus neutron fluence in dpa,

· the fatigue curve [ciad - N up to N = 1012 cycles,

It seems that the ABIT represent good tool for the measurement of the ultimate tensile strength, R_m. The extrapolation of obtained value to the operating temperature must be taken into account

For the evaluation of the fatigue curve k_b. - N up to N 10¹² cycles the equations were recommended, the true values of R_m^T , R_p^T_{0 2} and Z^r represent necessary inputs.

Based on this state of art of our knowledge we can stated that after 15 years of operation the changes of elastic- plastic and fracture-mechanic properties of irradiated stainless steel 08Ch18N1OT don't represent danger for reliable operation of the RPV internals for next 5 years.

Parallel to the preliminary re-assessment the preparatory works for the complex experimental testing of the active samples from the
Greifswald/>/> internals proceeded.

In the phase I of the project the following progress was reached:

· Hot cells are prepared for acceptance of the high active material samples from VVER internals; for test specimens machining; for mechanical and SCC testing; for microscopy of the material samples.

· The real austenite material from Greifswald RPV internals was studied in non-active state. It will be available the complex material characteristics for future evaluation of irradiation effect.

· The active samples from the
Greifswald/>/> internals have been satisfactorily transported to NRI.

· The ABIT method has been calibrated on massive samples. The ABIT calibration will be used for determination of VVER internals mechanical properties on NPP in operation.

Beneficiary Organisation Details

Czech State Office for Nuclear Safety:
http://www.sjub.cz

Project Aims

The purpose of the project was to establish a methodology for the experimental part of the work as well as to determine integration of the resulting database of properties into Czech regulatory procedures. Results gained through this project would help improve the quality of data supplied for the evaluation of safety assessment reports and inspection procedures in the context of the exercise of the licensing functions of the regulatory authority.

Project Results

The project was design to take advantage of a unique opportunity to study the changes in mechanical properties of actual internal VVER reactor components that would become available due to dismantling of Unit 1 at Greifswald, Germany. The project would result in a materials database.

The project was supposed to proceed in two phases. In the first phase, the requisite mechanical samples should have been procured from the Greifswald NPP and brought to the Czech Republic for storage prior to use. In a second phase, the following activities had to be performed:

• Determining tensile and fracture mechanics properties of material irradiated during operation (including base metal, weld metal, heat-affected zone and material of bolts).
• Determining other parameters necessary for re-assessment of internal structure according to particular methodology (crack growth rates; threshold stress intensity factor, fracture toughness of irradiated materials).
• Comparing high lead experimental reactor irradiated material properties with properties of the material irradiated in actual operational conditions (Greifswald NPP) and establish a correction factor.

The final project output was the development of relevant materials database.

The budget for this project was: 890 000 euro.

Further Project Results

Further information on the project could be sought from the beneficiary organisations.