Hence, these limits do not exist anymore in the application of up-to-date data logging systems. The application of modern BUS technologies paves the way for an economic and flexible measuring system without long cables. Handling and performance has been significantly improved for the modern systems. AREVA offers FAMOSi as a modern data logging system of that kind. It delivers the data base for the further fatigue monitoring and for the further assessment in a highly reliably way.

1.3 Data post processing and evaluation of cumulative usage factors (CUFs)

As it was mentioned before the evaluation of the recorded measurement data is done by application of a three staged assessment system. Only those analyses are applied that are required for the fatigue assessment of the according component. If the CUF obtained by application of the first analysis stage — the simplified fatigue estimation (SFE) — falls below the defined target value no further evaluation will be required. If it does not the second stage — fast fatigue evaluation (FFE) — will be activated and so on.

All data have to undergo a plausibility check before being further processed. The plausibility check enables a data control based on predefined limiting values. Erroneous data blocks originating for instance from electromagnetic pulses (e. g. switching operations within the main coolant pump) are detected by identification of physically questionable values of the temperatures or their gradients. These data sets are corrected to logical values. The correction is precisely recorded and can be reproduced at any time. Thus, the generated plausible data set constitutes the basis for the further data processing and assessment.

In fact, in the context of some severe external events, the assumption of continued availability of infrastructure required to administer emergency measures (for example roads and bridges) may not be valid. Under such situation, it is more effective to enhance the quality of the other levels of defence in depth. There is therefore, a need to define the scope of off-site emergency planning activities for advanced reactors, consistent with the ability of these reactor designs to meet enhanced safety objectives.

In some cases, such as the presence of a nearby airport, the consideration of the hazards may change risk zoning or eliminate a site from further consideration for an NPP, but most external hazards are either screened out from the necessity of being considered further or are taken into account in plant designing and siting. Risk zoning and siting is a matter for:

• The uncertainties of risk measures and influence to the public perception;

• Economic consideration (where power is needed, the availability of existing grid);

• Social and political factors;

• Topography affecting the dispersion of radio-nuclides through the atmosphere, rivers and ground-water;

• Political and safety consideration;

• Demographic characteristics;

• Hazards (natural and manmade).

Some IAEA Member States only address the risk to an individual member of the public, others have requirements to consider the potential aggregated effects to the population as a whole — societal risk.

Usually, off-site emergency measures are still seen as part of the Defence in Depth approach, which is mainly understood in deterministic sense, but to take full advantage of new reactor designs it should be moved towards a more probabilistic approach based on risk assessment with sensitivity and uncertainty analysis. The full benefit of innovative and evolutionary NPP requires the ability to licence without the need of an off-site Emergency Planning Zone.

In general, the desirability or possibility of reducing emergency response plans for accidents depends not only on the reactor type but also on a number of complex and intertwined factors including technical, societal, economical and cultural. The subject cannot be coupled directly and solely to the requirements for the external events but requires a separate consideration. Under the same subject also the risk-informed decision making related to the design basis accidents and severe accidents may be considered with the intent of moving away from somehow postulated risk zones and towards clearly calculated risk zones. Without such a change, related procedures and criteria, the issue of the emergency response plans cannot be resolved. In particular, in order to deal with external events and apply the risk-informed approach for plant design and siting, it is desirable to couple the PRA with analysis techniques of civil engineering.

Out of many semi analytical models, Figure 5 shows theoretical stability boundary for fluid — elastic instability obtained by (Lever & Weaver, 1986).

1. In order to attach a zero-point adjustment device firmly onto the CSB snubber lug, a zero-point adjustment plate and a connector on the gap control section were tightened. The remaining zero-point adjustment devices adhered to the CSB snubber lug as described above. Fig. 25 shows the zero-point adjustment device mounted on the CSB snubber lug.

2. Length measurement was done using the software program of the remote measurement computer. Measurements were taken five times. The average values were then used to set the zero-point data. When the setting of the zero-point data was complete, the zero — point data were saved and recorded.

3. The zero-point adjustment device was detached from the CSB snubber lug and a marking tool was attached to the CSB snubber lug. Digital probes were stained with a red stamping ink and their correct operation was confirmed using the software program of the remote measurement computer. The remaining digital probes were executed in the same way.

4. The marking tool was removed from the CSB snubber lug and the channel boxes in the CSB assembly and air compressor were then separated from the air hoses, electric power cords, and signal cables connected to the remote measurement computer. The disconnected cords and cables had to be arranged so that a disturbance did not result from the combination of the RV and the CSB assembly. Fig. 26 shows the marking tool assembly attached onto the CSB snubber lug.

Fig. 26. Installation of a marking tool

Starting from the calculated data developed in previous analyses (Pottorf et al., 2009; Mascari et al., 2009b; Mascari et al., 2011c) the target of this section is to give an expanded revised analyses, after a first review of the TRACE nodalization, of the TRACE V5 patch 2 code capability in predicting the primary/containment coupling phenomena typical of the MASLWR prototypical design.

The analysis of the OSU-MASLWR-001 calculated data shows that the TRACE code is able to qualitatively predict the primary/containment coupling phenomena characterizing the test. The blowdown phenomena, the refill of the core and the long term cooling, permitting of removing the decay power, are predicted by the code.

In particular, following the inadvertent middle ADS actuation, the blowdown of the primary system takes place. A subcooled blowdown, characterized by a fast RPV depressurization, is predicted by the code after the SOT.

When the differential pressure in the facility at the break location results in flashing, a two — phase blowdown, qualitatively predicted by the code, occurs. A decrease in depressurization rate of the primary system is then observed, in agreement with the experimental data.

When the PRZ pressure reaches saturation, single phase blowdown occurs and the depressurization rate increases again, in agreement with the experimental data. The RPV and HPC pressure versus code calculations are shown in figure 13.

In agreement with the experimental data, when the pressure difference between the RPV and the containment reaches a value less than 0.517 MPa, the high ADS valves are opened which equalizes their pressure.

When the pressure difference reaches a value less than 0.034 MPa, the sump recirculation valves are opened and the refill period begins. The refill phenomenon is predicted by the code. As in the experimental data, the refill period takes place for the higher relative coolant height in the HPC compared to the RPV.

Figure 14 shows the RPV level evolution experimentally detected during the test versus the calculated data. In agreement with the experimental data, the RPV water level never fell below the top of the core. Figure 15 shows the HPC level versus code calculation during the test. The qualitative behavior is well predicted by the TRACE code.

In agreement with the experimental data, during the saturated blowdown period the inlet and the outlet temperature of the core are equal each other assuming the saturation temperature value. A core reverse flow and a core coolant boiling off at saturation is predicted by the code. When the refill takes place, the core normal flow direction is restarted. Figure 16 shows the experimental data versus code calculation for outlet core temperature.

In agreement with the experimental data when the sump recirculation valves are opened the vapor produced in the core goes in the upper part of the facility and through the high ADS valve goes to the HPC where it is condensed. At this point through the sump recirculation line and the down comer the fluid goes to the core again (Mascari et al., 2011c, 2011d). Figure 17 shows the long term cooling flow path typical of the MASLWR design.

Figure 18 shows, by using the SNAP animation model capabilities, the fluid condition of facility, 976 s after the SOT, predicted by the TRACE code.

From a quantitative point of view the results of the calculated data show a general over prediction compared with the experimental data. It is thought that this could be due to a combination of selection of vent valve discharge coefficients and condensation models applied to the inside surface of the containment (Pottorf et al., 2009; Mascari et al., 2011c).

However, in order to quantitatively evaluate the capacity of the TRACE code to simulate OSU-MASLWR-001 primary/containment coupling phenomena, a qualification of the TRACE nodalization against an heat losses experimental characterization at different primary side temperature is necessary. Figures 19 and 20 show the behavior of PRZ pressure and outlet core temperature respectively by increasing the heat losses of the TRACE model (TRACE_HL). A general quantitative improvement of the calculated data has been showed by TRACE_HL calculation. Therefore in order to quantitatively evaluate the capability of the TRACE code to simulate the OSU-MASLWR phenomena, and therefore use the calculated data for the TRACE code assessment, is necessary a TRACE nodalization qualification against several facility operational characteristics like heat losses at different primary side temperatures and pressure drop at different primary mass flow rate. Currently the TRACE model qualification process is in progress considering the facility characterization that will be disclosed during the current IAEA ICSP.

Considering the importance of the containment/reactor vessel thermal hydraulic coupled behavior for the advanced integral-type LWR, in the IAEA ICSP framework a further test, simulating a loss of FW transient with subsequent ADS actuation and long term cooling, will be executed in the OSU-MASLWR facility (Woods & Mascari, 2009; Woods et al.; 2011).

6. Conclusions

The MASLWR is a small modular integral PWR relying on natural circulation during both steady-state and transient operation. During steady state condition the primary fluid, in single-phase natural circulation, removes the core power and transfers it to the secondary fluid through helical coil SG. In transient condition the core decay heat is removed through a passive primary/ containment coupling mitigation strategy based on natural circulation.

The experimental analysis of the primary/ containment coupling phenomena, characterizing the MASLWR design, has been developed in a first testing program at OSU-MASLWR
experimental integral test facility. In particular the OSU-MASLWR-001 test determined the pressure behavior of the RPV and containment following an inadvertent actuation of one middle ADS valve. The test successfully thermal hydraulically demonstrated the passive primary/containment coupling typical of the MASLWR design SBLOCA mitigation strategy.

In the last years the USNRC has developed the advanced best estimate thermal hydraulic system code TRACE in order to simulate operational transients, LOCA, other transients typical of the LWR and to model the thermal hydraulic phenomena taking place in the experimental facilities used to study the steady state and transient behavior of reactor systems. The validation and assessment of the TRACE code against the MASLWR passive primary/containment coupling mitigation strategy is a novel effort and it is the topic of this chapter. Since the qualification process of the OSU-MASLWR TRACE nodalization is still in progress, considering the facility characterization conducted in an IAEA ICSP on "Integral PWR Design Natural Circulation Flow Stability and Thermo-Hydraulic Coupling of Containment and Primary System during Accidents", the current results are preliminary and should not be used for the code assessment, but are able to show the TRACE capability to reproduce the thermal hydraulic phenomena typical of the MASLWR primary/containment coupling SBLOCA mitigation strategy.

The analysis of the OSU-MASLWR-001 calculated data shows that the TRACE code is able to qualitatively predict the primary/containment coupling phenomena characterizing the MASLWR design. The sub-cooled blowdown, two-phase blowdown and single phase blowdown, following the inadvertent middle ADS actuation, are qualitatively predicted by the code. The refill phenomenon is qualitatively predicted by the code as well. In general the results of the calculated data show an over prediction compared with the experimental data. It is thought that this could be due to a combination of selection of vent valve discharge coefficients and condensation models applied to the inside surface of the containment. In agreement with the experimental data, the RPV water level never fell below the top of the core. However, in order to quantitatively evaluate the capability of the TRACE code to simulate the OSU-MASLWR phenomena, and therefore use the calculated data for the TRACE code assessment, is necessary a TRACE nodalization qualification against several facility operational characteristics like pressure drop at different primary mass flow rates and heat losses at different primary side temperatures. Currently the TRACE model qualification process is in progress considering the facility characterization conducted during the IAEA ICSP.

7. Abbreviations

ADS

AP600/1000 CAREM CHF CL CPV EHRS ESBWR FW

Total energy; Gravitational force; Pressure; Temperature;

Phase velocity.

10.

The safe and reliable operation of steam generators is the essential pre-condition for the safe operation of the whole NPP, but also for all economical parameters at the unit. The steam generator has to be able to transfer the heat from the reactor in all operating or accidental regimes.

It is well-known that VVER-440 units have the horizontal steam generators with much higher capacity of cooling water in tank than in vertical steam generators, which are normally used in western NPPs. Undoubtedly; these horizontal steam generators are safer. On the other hand, due to horizontal design as well as in total 6 loops with additional welds and pumps cover a huge area, which is a limiting factor for containment construction of the unit. The exchange of steam generators is extremely difficult (in some reactor types almost impossible), therefore their optimal operation and clever maintenance (upgrades) is one of essential duty of NPP staff.

Based on operational experiences, the mitigation of damages and leak tightness defects in pipelines or collectors require much more time and money, than prevention measures. It is necessary to keep in mind the actual development in nuclear industry towards NPP lifetime prolongation and power increase (one of essential goals of European Commission 7FP — NULIFE). Fortunately, VVER-440 steam generators were designed with the huge power reserve (possible overloading of 20%). Beside several leakages in primary pipelines (0I6 mm), which can be in case of VVER-440 SGs relatively easy solved (blended), the corrosion deposits in feed water pipeline system occurred at many VVER-440 units [9]. The identified damages were caused mostly due to corrosion/ erosion processes attacking materials familiarly called "black steels" with insufficient resistance against corrosion.

Based on experience from Finland, also other countries including Slovakia changed the old feed water pipeline system. At this moment we would like to mention that the incident at the 2nd unit of NPP Paks (Hungary, 2003) connected to cleaning of fuel assemblies in special tank had the root causes in insufficient passivation of pipelines in SG after the steam water pipeline system exchange in 1997.

All steam generators at four VVER-440 units in Bohunice were gradually changed. At that time, there were two possibilities for the new feed water pipeline system. Out from two conceptions: Vitkovice a. s. design and OKB Gidropress design. The first solution was selected and lately improved to so called "Bohunice solution". Actually, experience from the last 10 years after upgrade was utilized.

A detailed description of VVER-440 steam generators delivered to NPP Bohunice is in [10­12]. The safety analyses were performed in 1977 by OKB Gidropress according to the Russian norms. The Russian designer and producer made the feed water pipelines (secondary side) from the carbon steel (GOST norm 20K and 22K). Water inlet pipeline was connected to the T-junction. From this point, 2 lines of the pipe with nozzles distributed the cooling secondary water in the space between primary pipes. Several problems having occurred in other NPPs were published in [13-16]. A disadvantage of such steam generators are difficult accessibility to the T-junction and next pipelines in the bundle.

An original concept of modular evolved control system, seamless and with gradual integration into the primary control system is proposed.

The aim of the application is to integrate the concepts of evolved control algorithms, portability of software modules, real time characteristics of the application.

The target systems are the large scale distributed control systems with optimum granularity architecture.

The first part of the life cycle phases of the new control system, from conception to validation stage, the new control system lives "hiding in the shadow" of the control system it will replace, and after validation the old system will be replaced by the new one.

The identification, modeling, control and validation stages of the life cycle of the system, will be done on-line (the new system uses a real image of the I/O process data), without affecting the existing control system.

Because of high level of interconnectivity between system components, it is necessary to provide the highest independence between communication modules on one-hand and the control modules on the other hand. In order to obtain high ability of integration, the communication modules have to cover the widest possible area of industrial communication interfaces and protocols.

One item of the application is to offer a unified API of extended generality and extendibility in order to unify access and information retrieval from various wireless and wired technology and communication interfaces (RS 232, RS 485, fieldbus: Profibus / Interbus, Ethernet IP, TCP/IP, etc).

Applications could properly adapt to changes in the network connections.

The design and implementation of a solution to hide the embedded communication network problems from the application system programmers is included.

One of the main objectives of the application is to supply an integrated solution of systems, which should support all the phases of the life cycle: modeling, simulation, development and implementation.

For parameter tuning, for validation and also for embedding a large number of industrial communication protocols, multi-disciplinary simulation environments are developed which generate instruments for control, I/O data consistency check, and defect detection.

In the end, real-time advanced control applications are developed, with seamless and gradual integration into the existing distributed control system.

A software package for evolved control includes a method based on fuzzy model predictive control.

By using the basic concept of decomposition-coordination in a large-scale system theory, the fuzzy model predictive controller design can be accomplished through a two-layer iterative design process.

The design is decomposed into the derivation of local controllers. The subsystems regulated by those local controllers will be coordinated to derive a globally optimal control policy.

In order to provide the real-time characteristic, we choose a multitasking environment for the application (WINDOWS Operating System).

From structural point of view we propose a Client / Server architecture for fuzzy Controller (FC) (Andone et al., 2006):

Client — is a Windows application representing the implementation of the graphical user interface (GUI). The Client enables the operator to control the system in two modes: manual/automatic, to monitor the system response, etc. The Client has also the ability to connect and communicate with the Server application.

Server — is an ActiveX EXE application containing the implementation of the Fuzzy Controller (FC) kernel.

The Server includes a collection of objects, these objects cover the tasks of both data processing and the communication between dedicated applications for input and output data.

The Client application will have a thread pool architecture.

The Server application will have a real multithreading architecture (each active object having assigned its own execution thread).

The Server have also a multi-layer structure: at the higher level are implemented upper FC and the communication classes (using different transmission mechanisms — DDE, OPC, HLI, ActiveX, Winsocket, Pipes), at the lower level are implemented the controllers for the subsystems corresponding to the low level FC.

The Server’s application as real multithreading architecture, provides the FC Kernel the real­time response characteristic, required for the industrial process control.

4. Conclusions

Control of SG water level strongly affects nuclear power plant availability.

The control task is difficult for a number of reasons, the most important among them being the nonlinear plant dynamics and the non-minimum phase plant characteristics.

There has been a special interest in this problem during low power transients because of the dominant reverse thermal dynamic effects known as shrink and swell.

The SG level control problem was viewed as a single input/single output control problem with the feed-water as the manipulated variable, the level as the controlled variable and the turbine steam demand as disturbance.

It has been shown that in the case of nonlinear processes, the approach using fuzzy predictive control gives very promising results.

The process non-linearity was addressed by scheduling the model (and the controller) with the power level.

The SG system is modeled by Takagi-Sugeno’s fuzzy modeling methodology, where the system output is estimated based on gradient. The complex shrink and swell phenomena associated with the SG water level are well captured by the model.

The predictive controller based on fuzzy model is designed in a hierarchical control design.

An original concept of modular evolved control system, seamless and gradual integration into the existing distributed control system is proposed in the chapter.

A unified API of extended generality and extendibility in order to unify access and information retrieval from various wireless and wired technology and communication interfaces is developed in order to ensure independence between communication and control modules of the designed systems.

A Client / Server architecture for evolved controller that runs on the Windows

environment, with real-time characteristics is proposed.

5. Acknowledgment

Parts of this chapter are reprinted from Hossu, D. Fagarasan, I., Hossu, A., Iliescu, S. St.,

Evolved fuzzy control system for a steam generator, Int. Journal of Computers, Communications

and Control, IJCCC, ISSN 1841-9836, E-ISSN 1841-9844, Vol. V (2010), No.2, pp. 179 — 192.

Table 6, an extension of period beyond 1993 (Pettigrew et al., 1973) presents a summary of salient features of the experimental tests performed on the three possible tube arrangements (triangular, normal square, and rotated square).

Multi-objective decision making builds on previous multi-objective (sometimes called multi­attribute) valuation of the alternatives. Because the different ways to solve the problem tend to be mutually exclusive, the selection of the "best" option requires the formulation of trade­offs among the different attributes used to evaluate the performance of the several possible alternatives. Such trade-offs require a multi-objective analysis (see above) in order to assess and compare the relative merits of the alternatives. In practice, a multi-objective analysis usually does not yield a single optimal alternative. Therefore, the choice of the "best" solution requires that the decision maker’s preferences and value trade-offs among conflicting objectives be clearly articulated and made explicit in the selection process. A vast number of publications on multi-attribute decision making is available from which one can extract useful information and guidance on how to perform such decision modelling. The following selection may serve as an introductory reading to the comprehensive overview of approaches, methods and tools for different multi-objective decision applications (Bohanec & Rajkovic, 1999; Bohanec, 2003; Munda et. al., 1995).

3. Radioactive waste disposal

3.1 Perception of radioactive waste disposal issues

The recent international perspective can be found in the report "Resource or waste? The politics surrounding the management of spent nuclear fuel in Finland, Germany, Russia and Japan" (SKB, 2011). A clear historical divide can be discerned between countries that decided to reprocess spent nuclear fuel and those that chose final disposal. Three of the countries mentioned — Japan and Russia and, in an earlier phase, Germany, have considered spent nuclear fuel as a resource rather than as waste, and for that reason invested in reprocessing. The report provides an account of how and why these countries chose different alternatives; why, despite a common basic approach, they gradually came to aim at completely different strategies and methods for spent nuclear fuel management. Today Germany has totally abandoned its previous reprocessing strategy, Russia has maintained its strategy, but also steered certain operations toward direct disposal, and Japan has recently completed a major industrial reprocessing facility. The issue of final disposal is, however, far from solved in Germany and Japan. In order to understand why different countries have chosen one alternative over another, and how a strategy changed over time, the authors chose to elaborate on eight key dimensions. Five of these relate to nuclear power issues, such as whether or not a country produces nuclear weapons, has an expanding or stagnating nuclear power sector, weak or strong competence in the field of nuclear energy, good or poor prerequisites for a final repository, and whether or not it has domestic uranium resources. Three other dimensions cover political characteristics, i. e. whether or not the country had or has a strong or weak anti-nuclear movement, whether it is a democracy or a dictatorship, and whether or not it is characterized by strong or weak local political power. The latter aspect is seen as essential to issues of local acceptance of a spent nuclear fuel repository. The reasons behind different choices appear to be the military use of spent nuclear fuel and the absence of democratic discussion (Russia), consensual political decision-making (Finland), and situations of strong political opposition and local disputes (Germany and Japan).

In the project "Nuclear waste: From an Energy Resource to a Disposal Problem" (SKB, 2011) Jonas Anshelm analyzed the nuclear waste debate since the 1950s, including issues of risk, responsibility, design of a final repository and safety of the technology. The author points to the importance of elucidating the different kinds of answers that have been given concerning these issues in different time periods. The challenge is to understand how changing technological, political, economic and scientific circumstances have influenced perceptions and debates. Such clarification can broaden the perspective and facilitate an understanding of the complexity of the issue. The project observes shifts in meaning and public opinion changes regarding central aspects on the nature of nuclear waste — as a resource or as a waste, and the characteristics of the waste — as well as of its associated risks. Likewise, issues of who has responsibility for the final repository, what should be considered scientific facts concerning bedrock characteristics, and the sustainability of the technological solutions, have been subjected to controversy throughout the period. It is striking, Anshelm notes, that central actors have been both utterly confident in their opinions and able to assume totally different points of view in new situations. This characterization applies to both proponents and opponents of nuclear power. In summary, this contribution illustrates that what is perceived to be true, valid, correct, morally right, and rational with respect to the debated issue has recurrently been subject to renegotiation and change during the past half-century. This has resulted in a number of serious conflicts since the 1970s. The issue has currently reached a level of stabilization and does not exemplify a strong national or local controversy. It is, however, reasonable to assume that current views on what is true and right regarding the nuclear waste issue — on which there is some consensus today — will, in the future, also be subjected to renegotiations in the light of scientific, technological, economic and political reorientations. This already appears to have been triggered in a number of countries, e. g. Germany, Japan, Slovenia, by the consequences from the damaged NPP Fukushima I after the quake and tsunami in March 2011. It could be viewed that this accident encouraged the German government to announce that it will bring forward the closure of its nuclear power stations to 2022, 14 years earlier than originally planned, while Japan considers a review of plans for construction of new NPPs, just like Slovenia in its new National Energy Programme currently under debate.

2.2 Radiochemical separation

The methods for separating, collecting, and detecting radionuclides are similar to ordinary analytical procedures and employ many of the chemical and physical principles that apply to their nonradioactive isotopes. One of the differences is interesting from the viewpoint of methodology. Substance separation in analytical chemistry in the majority of cases is not an end in itself. In radiochemistry, separation is most often an end in itself, for example, when a radionuclide is purified of other radioactive elements (Zolotov, 2005). Techniques used for separation include co-precipitation, liquid-liquid extraction, ion exchange and extraction chromatography. In some cases, two or more of these techniques are combined.

In order to account for the inevitable loss of the sample during separation, a specific isotope or tracer is added to the sample. A tracer represents the addition to an aliquot of sample a known quantity of a radioactive isotope that is different from that of the isotope of interest but expected to behave in the same way. Sample results are normally corrected based on tracer recovery. The percent of tracer lost in the chemical processes is equal to the percent of sample lost, assuming the tracer is homogeneously mixed with the sample and is brought into chemical equilibrium with the sample. Radiochemical analysis frequently requires the radiochemist to separate and determine radionuclides that are present at extremely small quantities. The amount can be in the picomole range or less, at concentrations in the order of 10-15 to 10-11 molar (United States Environmental Protection Agency, 2004). The use of a material that is different in isotopic make-up to the analyte and that raises the effective concentration of the material to the macro level is referred to as a carrier, a substance that has a similar crystalline structure that can incorporate the desired element.

Radiochemical waste characterization is the identification of radionuclides contained in a package of nuclear waste and the determination of their concentration. The problem the waste producers have to cope with comes from the fact that those nuclides which are mainly (pure) |3- or а-emitters cannot be measured by direct methods such as у-scanning. In the waste packages produced by a nuclear power reactor the radionuclides may be originated as fission products from the nuclear fuel, activation products and transmutation nuclides, Table 1.

Identification of these nuclides requires methods that, in general, involve analyses of waste samples using complex chemical analysis to separate the various radionuclides for measurement. Among the various proposed methods there are those who seek the identification of a radionuclide isolated or those seeking to identify by simultaneous determination two or more radionuclides in the same analysis.

The main constraint for a new protocol is to obtain a high recovery yield, a high-energy resolution and low interferences of other radionuclides. Thus, it is necessary to develop accurate and reliable methods for the determination of radionuclides in the low and intermediate radioactive samples. A simultaneous determination procedure was developed for the separation of Pu isotopes, 241Am, 242Cm, 244Cm, 89Sr and 90Sr using precipitation by oxalate, ion exchange resin, extraction of plutonium by TTA (thenoyltrifluoro acetone/benzene) and Sr by precipitation techniques. This method was applied for determination of these radionuclides in the grass, collected near Munich after the fallout from the nuclear accident at Chernobyl (Bunzl & Kracke, 1990). In another case, Pu, Am and Cm were determined by extraction chromatography using an organophosphorus compound immobilized on an inert support commercially available under the name TRU Resin (for Transuranium specific) from Eichrom Technologies, Inc. This method was used in samples from nuclear power plants such as spent ion exchange resins and evaporator concentrates (Rodriguez et al., 1997). Besides, combined procedure was used for the determination of 90Sr, 241Am and Pu isotopes by anion exchange for Pu isotopes analysis, the selective method for Sr isolation based on extraction chromatography using Sr Resin and the TRU Resin for separation of Am (Moreno et al., 1997). In the radiological characterization of low — and intermediate-level radioactive wastes the separation of Pu isotopes, 241Am, 237Np and 90Sr was performed by anion-exchange chromatography, extraction chromatography, using TRU and Sr Resin, and precipitation techniques (Tavcar et al., 2007).