Surveillance embrittlement

Server, W., Eott, R., and Rosinski, S. (2004) Assessment of U.S. Embrittlement Trend Equations Considering the Latest Available Surveillance Data, PVP-Vol. 490, Storage Tank Integrity and Materials Evaluation, Jule 25-29, San-Diego, California, USA, 19-24... 

A European code of practice for reconstitution of irradiated Charpy specimens (used to monitor the embrittlement behaviour of the RPV) was developed within FP-4 project RESQUE. This should allow to re-use surveillance specimens, which were tested destructively in order to determine the fracture toughness. 

Key words pressurized water reactor (PWR), reactor pressure vessel (RPV), surveillance database, radiation embrittlement. 

In this chapter, after giving an overview of the embrittlement of Western pressurized water reactor (PWR) reactor pressure vessel (RPV) beltline materials, together with the characteristics of PWR RPVs, such as their general specification, core region materials and the effects of variables on embrittlement, the surveillance database obtained from US, French and Japanese nuclear power plants (NPPs) and those from other countries is presented based on open literature. The surveillance program of each country is also briefiy described. Trends of surveillance data which will be obtained in the near future are described. The possibility of new data from reconstituted and miniature specimen techniques is described. 

ASTM E185-82 also provides the option of including correlation monitor material in the capsules. Correlation monitor material provides an independent check on the irradiation conditions because other specimens of the same correlation monitor material have been irradiated in other reactors and its response to irradiation (embrittlement characteristics) is well documented. An example of a common correlation monitor is Heavy Section Steel Technology (HSST) Plate 02 (an SA533 Grade B, Class 1 material). Correlation monitor material is also included in the surveillance program to provide a means of validating neutron fluence estimates for the surveillance capsules. The data obtained from correlation monitors should... 

Incorporating these corrections and updates, and adding additional surveillance data from recent capsule testing, the database and NRC embrittlement shift model in NUREG/CR-6551 were updated again in July 2000. Separately, the Electric Power Research Institute (EPRI) and the ASTM E10.02 subcommittee evaluated the database and derived a slightly simpler embrittlement correlation model published as E 900-02 in 2002. 

A significant amount of new data became available in 2003 when capsules from the BWR Supplemental Surveillance Program (SSP) were tested. Because low-flux BWR shift data were significantly under-predicted by the existing shift models, additional refinement of the NRC model was performed. Furthermore, the Eason database was updated for new PWR capsule test data (about 140 shifts) and then used to derive the embrittlement correlation known as EONY (after the developers Eason, Odette, Nanstad and Yamamoto) (Eason et al., 2007) which was codified in the Alternative PTS Rule, 10 CFR 50.61a (10 CFR, 2010). 

Several industry databases of surveillance test data have been maintained over the years. For example, during the 1990s EPRI maintained the Power Reactor Embrittlement Program (PREP) database (EPRI, 1996), which was developed under joint EPRI and CRIEPI (Central Research Institute of the Electric Power Industry) sponsorship. Later, EPRI sponsored the RPVDATA database program (EPRI, 2000) as a successor to PREP. The industry databases have faced greater sustainability issues than the NRC databases because of limited funding. 

Tomimatsu M, Asada S, Namatame H, Ohtani M and Watada M (1996), Evaluation of RPV steel surveillance program in Japanese PWR radiation embrittlement, prediction . Proceedings of P" International Symposium on REACTOR DOSIMETRY,Ahde m] m H A, O Hondt P and Osmera B (eds), Prague, Czech Republic, 2-6 September, 1996, World Scientific, Singapore. 

Abstract This chapter describes the embrittlement processes in WWER reactor pressure vessel (RPV) materials during operation - radiation embrittlement and thermal ageing. Current trend curves for both types of WWER RPV materials are given and explained. Surveillance specimen programmes are shown, as their results are used for RPV integrity and lifetime evaluation. Finally, anneahng of the RPV is proposed as the most efficient mitigation measure. 

For calculation of 50 % (median) curves of AT i(F, t) according to Eq. 5.5, the value a = 0 should be taken, both for base and weld metal. This type of dependence is used for direct determination of radiation embrittlement parameters based on results of RPV-specilic surveillance specimen tests. Typical trend curves calculated according to the Eq. 5.4 for base metal and weld metal are shown in Figs 5.2 and 5.3. It can be seen that mean curves provide a good description of the distribution of test results from surveillance specimen programmes. 

The lead factor in surveillance specimens for upper floor assemblies is lower than 1 and therefore such results cannot be used for prediction of irradiation embrittlement of RPVs. 

Methods for determination of fracture toughness for calculation of strength and lifetime of WWER-1000 RPVs on the base of surveillance specimen test results, RD EO 1.1.2.09.0789-2009. Appendix G. Procedure for determination of irradiation embrittlement of WWER-1000 RPV materials as a function of... 

The prediction of embrittlement shift in transition temperature is generally based on correlations of measured surveillance CVN transition temperature shifts with specific chemistry variables (generally Cu, Ni, P, Mn and Si) and fluence for the materials of interest. 

It is widely known that radiation embrittlement behaviour of reactor pressure vessel (RPV) steels depends on various parameters such as material composition, neutron flux and irradiation temperature. Sound understanding and modelling of embrittlement mechanisms require systematic knowledge of effects of individual parameters and their synthesis on microstructural development and then mechanical properties. Most such knowledge has been obtained from single-parameter experiments using test reactor irradiation. This is because test reactor irradiation allows researchers to obtain mechanical property data together with microstructural data on materials with well-controlled chemical compositions under well-controlled irradiation conditions such as flux and temperature. Surveillance data in commercial power reactors are non-systematic in this context and relevant microstructural data are very scarce. 

One of the merits of test reactor irradiation is the higher dose irradiation at higher flux compared to irradiation at vessels or surveillance capsules in commercial power reactors. As of the end of 2013, the maximum fast neutron fluence of surveillance data was around 9 x lO n/cm E> MeV) while the maximum fluence at the inner surface of a RPV may exceed 1 x 10 °n/cm E > IMeV) for a 60-year operation in a pressurized water reactor (PWR). Test reactor irradiation is the only way to obtain experimental data for embrittlement behaviour at fluences relevant to 60 years or longer operation, together with an understanding of neutron flux effects on embrittlement behaviour. 

The first two mechanisms contribute to embrittlement by increasing the hardness of steel as illustrated in Fig. 9.41. The third mechanism induces embrittlement without hardening. The latter mechanism is not necessarily found in all RPV steels under operating conditions. Indeed, for Mn-Mo-Ni steels irradiated in surveillance schemes in western light-water reactors (LWRs), the observed embrittlement is associated with the first two mechanisms (see Fig. 9.41), i.e. the total shift in the ductile to brittle transition, as measured at the Charpy 41J level, is... 

J.C. Van Dnysen, J. Bonrgoin, P. Moser and C. Janot, Study of the neutron damage in a PWR pressure vessel steel after long-term irradiations in the Chooz A Reactor Surveillance Programme , Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels An international review (4th Volume), ASTM STP1170, L.E. Steele, ed., American Society for Testing and Materials, Philadelphia, PA, 1993,132-138. 

---