1.1 Structure of the radiation factor in object "Shelter"

The unique biota inside object "Shelter" during its existence was appeared. It is one more physical carrier of ionizing radiation. High humidity and positive temperatures inside "Shelter" even in winter are responsible for the formation of homogenecity thermostatic conditions favorable for the development of microorganism biocenoses. Authors [Pazuhin E. M., Krasnov V. A. & Lagunenko A. S. 2004] have shown that the fuel-containing materials are dissolved under the action of microorganisms. As a result, new compounds of radionuclides with organic matter appeared and they are potentially more mobile and more ecologically dangerous. It was found out the microorganisms on the irradiated nuclear fuel from "Shelter" by the method of electronic microscopy. They are capability to develop on the irradiated nuclear fuel (dioxide of uranium) as a source of mineral substances and to cause the characteristic damage of a surface of nuclear fuel. In Figures 1 and 2 the characteristic damages of a surface of the irradiated nuclear fuel is found.

The size of fuel particles in the sample is in the range of 1 to 100 mcm, which corresponds to the grain size of the irradiated fuel is not oxidized. The main source of radioactivity was radionuclide 137Cs which moved through the aerosol pathway. Practically the entire aerosol 137Cs is detected in the material of organic origin fuel particles coated this organic material. Presence of radiating fields of different intensity promotes strengthening of mutational process and selection of radio-resistant microorganisms with new properties was investigated.

Dust. Using special equipment as called impactor received gas-air mixture probe samples inside object "Shelter" was determinated. This instrument allows you to separately collect particles of different aerodynamic diameters (from 9.4 microns up to 0.1 microns) of the aerosol complex structure with an uneven distribution of radioactivity. The rate of pumping of air was 70 ± 4 l/sec. The aerosol particles in the electron microscope REM-100U were evaluated. The most typical kinds of aerosol particles are represented in the Figures 3 — 6.

The particles of the correct spherical or oval-oblong, which are easily identified (spores, cocci, etc.) [Gusev M. V. & Mineeva L. A. 1992] and particles of irregular forms seen in presented figures. It is shown by electron microprobe elemental analysis that the major part of these particles consists of organic matter and it is of biological origined [V. B. Rybalka, G. F. Smirnova & G. I.Petelign, 2005].

Thus, there are a large number of particles consisting of organic matter in the investigated samples of aerosols originated from "Shelter" premises. A significant number of these particles are identified as the cells of microorganisms and spores.

Sub-micron hot particles (smHP) There is high-dispersive hot particles, which needs a special attention. Such kind of particles is a product of spontaneous dust productivity phenomenon which means dust generation from surfaces of irradiated nuclear fuel and LFCM surfaces. Such a phenomenon was discovered experimentally [Baryakhtar V. G., Gonchar V. V. & Zhidkov A. V. 1997] and for irradiated fuel was confirmed in [Walker C. 2000] later. The smHP grade distribution and possible physical mechanisms responsible for it was later identified in [Baryakhtar V., V. Gonchar & Zhidkov A. 2002]. Typical particle grade is 150 nm for fuel HP and near 50 nm for LFCM particles; all the particles have a complicated internal structure [Baryakhtar V., Gonchar V. & Zhidkov A. 2002]. Their radionuclide composition and specific radioactivity does correspond to those for FCM [Baryakhtar V. G., Gonchar V. V. & Zhidkov A. V. 1997]. Such kind of HP practically cannot be trapped by standard respirators, which usually at personnel’s disposal. Annual estimated activity generated in a form of such kind aerosol does equal to a few tens kilograms of irradiated fuel [Baryakhtar V. G., Gonchar V. V. & Zhidkov A. V. 1997].

Such sub-micron HP (Figures 7, 8) to be considered as the most dangerous radiation — hazardous agent regarding to a few reasons: their behaviour in biological liquids is similar to those for the particles in true liquid solutions, smHP does not need solubility of fuel matrix for penetrating all natural biological barriers originated from cell membranes. Sub­micron HP aerosol provides near 80% of total inhalation dose and to be an agent determining effective dose formation for "Shelter" object personnel [Bondarenko O. A., Aryasov P. B. & Melnichuk D. V. 2001]. Existing national regulatory documents on radiation safety does not establish a tolerable concentration of such kind aerosols in the "Shelter" object atmosphere because any attempt to classify them (in accordance to accepted classification) turned out to be doubtful. There are, however, explanations in regulatory document СПРБ-ОУ (in Ukrainian) (Appendix 1), which prescripts what should be done when real aerosols characteristics differs from the typical ones. According to that document, when planned activity stipulates thermal or chemical impact on FCM congestions or heavily contaminated "Shelter" object elements and when revealing in aerosols of non-oxide uranium or TUE chemical compositions one should establish the tolerable concentration for a- and P-emitters in air basing on results of additional special investigation.

Water of the lower marks and technological waters Shelter. It was established that concentration of у-emitting radionuclides reaches 3.8.1011 Bq/l, while the concentration of uranium measured up to 0.3 g/l in several thousand cubic meters of "block" and the technological water in premises of lower marks. These water accumulations influence the state of nuclear security of the object "Shelter". They may lead to a change in breeding properties of system "FBM + fragments of the core + water" and the emergence of emissions of short-lived radioisotopes such as iodine. This radionuclide is known to cause the spectrum of thyroid diseases, including cancer.

The concentration of radioactivity elements occurs in the silts as the drying pools during the summer-autumn period. Thus silts in "Shelter" represent a real risk as a source of radioactive aerosols. Different kind of microorganisms are found in water accumulations at zero marks of "Shelter" — bubbler pool (Figures 9 — 12) and on the surface of the walls (Figures 13 — 16) [Rybalka V. B., Rybalko S. I. & Zimin Yu. I. 2001, Zhdanova N. N., Zaharchenko V. A. & Tugaj T. I. 2005, Rybalka V. B., Smirnova G. F. & Petelin G. I. 2005, Petelin G. I., Zimin Yu. I., Tepikin V. E. 2003, Rybalka V., Klechkovskaja E., Serbinovich V. 2001]. The experimental investigation of water samples of the Shelter are presented in Figures 9 — 13.

As follows from figures 9 and 10, in the waters of the Shelter are present microbial community in various forms and sizes. In some samples well-reviewed spores size of about 0,1 micron. There is a large number of different units and formations. The biomass of the samples has a very high level of specific activity (137Cs to 3,9 x1010 Bq/ m3; 90Sr to 7, 9 x109 Bq/m3, 238+239+240 Pu to 1,1 X105 Bq/m3).

Microscopic siza fungi widely exist in the microbiota of the Shelter together with bacteria represented. It is shown that the defeat of microscopic fungi growing in areas with low levels of contamination (from one to 100 mR/hr) [Zhdanova N. N., Zakharchenko V. A. & Tugay T. I. 2005]. The life-cycle reduction, increased radioresistance, increasing the frequency of occurrence of positive radiotropic reactions and radiostimulation, high photosensitivity, that correlated with radiotropic response are typical for fungi-extermophyles isolated from such premises. In these premises increasing the risk of fungal biodegradation, increasing the probability of development of active agents of onychomycosis, skin lesions, lung infections, otitis, invasive fungal infections, as well as the selection of radio resistant organisms with unpredictable invasive characteristics. Lack in human natural immunity to these microbes, resistance of new cultures to the action of traditional medicines, high-speed distribution of microbes (biomass doubling from 15 minutes to 2 hours), the possibility of "transfer" the genetic information with hazardous properties to other bacteria, protozoa, algae, fungi and higher organisms can provide a very serious threat for people. It is known that in the light roof of sarcophagus has a large number of defects and inside the "Shelter" are circulating constant vertical wind currents. So it is quite realistic assumption that some of these microorganisms, especially in the form of spores, can potentially overcome the filters and act to the respiratory tract and lungs of people working inside this object at one side and at another side may carry away outside "Shelter".

For the experimental measurements, several specimens containing corrosion products were taken from different parts of all of 4 NPP Bohunice units. In the first step, corrosion process at the steam generators was studied. The corrosion layers were separated by scraping the rust off the surface and the powder samples were studied by transmission Mossbauer spectroscopy. It should be noted that the gamma spectroscopic measurements gave no evidence of the presence of low-energy gamma radiation emitted from the samples. Later, the corrosion products were collected also from different parts of secondary circuit components and several filter deposits were analysed as well.

The room temperature Mossabuer study was performed on two different steam generator materials using conventional transmission Mossbauer spectrometer with the source 57Co in Rh matrix. The spectra were fitted using NORMOS program.

The original STN 12022 material used at the 4th (SG46) over 13 years was compared to STN 17247 steel used at the 3rd unit (SG35) for about 5 years (1994-1998). The chemical compositions of both materials are shown in Table 1.

Samples of corrosion products scrapped from different parts of the steam generators SG 35 and SG46 were analysed. The scrapped corrosion particles were homogenised by granulation and sieved through a of 50pm wire sieve.

A simplified heat flow diagram of RBMK reactor is provided in Figure 1 [1]. The reactor cooling water, as it passes through the core, is subjected to boiling in the fuel channels (2) and is partially evaporated. The steam-water mixture then continues to the large drum separator (3), the elevation of which is greater than that of the reactor. The water settles there, while the steam proceeds to the turbines (5). The remaining steam beyond the turbines is condensed in the condenser (7), and the condensate is supplied by the condensate pumps (8) into the deaerator (9). Deaerated water is returned by the feed pump (10) to the drum separator (4). The coolant mixture is returned by the main circulation pumps (11) to the core, where a part of it is again converted to steam. The reactor power is controlled using control rods (3).

This fundamental heat cycle is identical to the Boiling Water Reactor cycle, extensively used throughout the world, and is analogous to the cycle of thermal generating stations. However, compared to BWRs used in Western power plants, the RBMK-1500 and RBMK — 1000 have a number of unique features. The comparison of most important parameters of the reactor is presented in Table 1. As it is seen from the presented table, the values of specific power per fuel quantity are very similar for all reactors. The value of power per fuel rod length is the highest for RBMK-1500 reactor. To reach such high value, additional specifically designed spacers, which operate like turbulence enhancers to improve the heat transfer characteristics, are mounted in the fuel assemblies of RBMK-1500. Specific power per core volume in RBMK-1500 is higher than in RBMK-1000 reactor, but in BWR-type reactors this characteristic is approximately 10 times higher.

In RBMK-type reactors a part of Reactor Cooling System (RCS) above the reactor core is located outside the leaktight compartments. In the RBMK reactors design, these compartments are called Accident Localization System (ALS). This is different from the typical PWR or BWR plants, which have full containment [1]. The Drum Separators (DS) and a part of downcomers are contained in the DS compartments, which are connected to the reactor hall. Such compartments are not as strong as the leaktight compartments of ALS.

This model assumes no relative velocity between the liquid velocity U1 and the gas velocity Ug . Slip S between the two-phases is:

S = 1: Uh = Ug = U{,

jg

jg + ji

where Uh is the homogeneous velocity, Ug is the gas phase velocity, Ut is the liquid phase velocity, sg is the homogenous void fraction, jg is superficial gas velocity and jt is the superficial liquid velocity.

3.1.1 The time perspective of nuclear waste

The most discussed aspect of nuclear waste is its longevity. Previously nuclear waste was the only issue for social decision-making that was widely discussed in very long time perspectives. Today climate change is discussed in such long time perspectives, and we also have a general discussion on sustainable development that does not have any time limits (Hansson, 2011). Hansson further says that discussions on decisions related to very long time perspectives include the issue of how to evaluate outcomes in the future. For example, is the value of a human life similar or dissimilar if it relates to assessing a final repository in e. g. 10,000 years hence, or in our time? And how should uncertain outcomes be evaluated? We seldom know about the consequences, in a hundred year perspective, of a decision taken today. This uncertainty has often resulted in not caring for the long-term consequences of the actions. The nuclear waste issue has become a pioneering case in the sense that uncertainties have not hindered us from considering long-term consequences seriously. Hansson’s concluding observations are that it is not the uncertainty per se that has resulted in the high attention and controversy regarding future effects of a nuclear waste repository, but rather the combination of certainty in specific areas (e. g. radioactive decay over time, etc.) and uncertainty in other areas (e. g. future generations’ knowledge, intentions, etc.). Finally Hansson notes that the International Climate Panel (IPCC) focuses on a time perspective of around 100 years and utilizes a kind of "trimmed discounting" in the work. He concludes that this is rather unprincipled reasoning, and suggests that much would be achieved by approaching the climate change issues in a way similar to that of nuclear waste.

All reagents used were analytical grade. The detection of radioactive 63Ni was carried out by Liquid Scintillation Counting (LSC), using the Quantulus 1220 spectrometer, the vials used were the 20 mL polyethylene and the scintillation cocktail was the Optiphase Hisafe 3, all from PerKinElmer Inc. (PerkinElmer Inc., Finland). The column materials used in the analysis were Ni Resin in pre-packed 2 mL columns, 100-150 g particle size, an extraction chromatographic material available from Eichrom Technologies (USA) and the anion exchange resin Dowex 1×8, Cl — form, from Sigma-Aldrich Chemical Co., (USA). 59Ni was analyzed using Ultra-LEGe Detector (GUL) with a cryostat window of beryllium low energy y-detector containing an active area of 100 mm2, efficiency 5.9 keV for 55Fe with a resolution of 160 eV in terms of FWHM, from Canberra (USA). The recovery was obtained analyzing stable nickel by ICP-AES.

The second category of core damage (see Figure 5), accidents with loss of the general structure integrity of the reactor and ALS, can potentially be due to the possibility of multiple ruptures of fuel channels at high pressure in RCS. The structural integrity of RBMK-1500 reactor depends on the integrity of the reactor cavity, which with a conservative strength margin was designed for conditions of an anticipated accident caused by the rupture of a single channel (in the nominal operating regime of a reactor). During operation of all nuclear power plants with RBMK reactors, there were three cases of fuel channel rupture, but the neighboring channels were not damaged [4]. This shows that the neighboring fuel channel — graphite cells have sufficient strength and elasticity and that the load caused by the rupture of a single channel is small. The accidents, leading to the complete reactor core damage, with loss of the structural integrity of the reactor can be divided into two groups [9]:

• accidents when heat-up of the reactor core occurs during the reactor operation or within the first seconds after the reactor shutdown, when decay heat is high (2.1, see Fig. 5);

• accidents when heat-up of the reactor core occurs after the reactor shutdown (2.2, see Fig. 5).

The first group of accidents (2.1) includes accidents when the heat-up of the reactor core occurs in the beginning of the accident (a few seconds after reactor scram activation) when decay heat in the core is high and the temperatures of fuel cladding and FC walls in the group of channels can reach the dangerous limits. An example of such accidents is the group of accidents, when the local flow stagnation occurs in the group of fuel channel during the LOCA. This situation is possible in the partial break case [12, 22]. It was mentioned earlier that partial rupture of lower water pipe could lead to the flow stagnation in the affected fuel channel. In the case of GDH partial rupture, the flow stagnation can occur in all fuel channels connected to this affected GDH. Under adverse conditions, the partial break of MCP pressure header can cause flow stagnation in the fuel channels of one affected loop. Since there is no sufficient time for actions of the operator in this case, the short-term accident management measures (automatic actuation of safety systems) are necessary. New reactor protection against coolant flow rate decrease through GDH generated signal for early ECCS activation is implemented in RBMK-1500 in this case. This short-term measure leads to the disturbance of the coolant flow rate stagnation in the group of fuel channels [12] and all parameters of RCS, thus, the reactor remains within safe limits.

Other examples of the first group of accidents (2.1) can also be the initiating events, namely:

• Anticipated Transients Without reactor Scram (ATWS);

• GDH blockage;

• Loss of natural circulation due to a sharp decrease of pressure in the RCS.

The analysis of ATWS (performed for the RBMK-1500 in 1996) demonstrated [3] the lack of inherent safety features in the RBMK design. The power is not reduced by means of inherent physical processes such as steam generation. The reactivity loss due to the fuel temperature rise (Doppler Effect) is not sufficient. The consequences of the accident for RBMK-1500 reactor, during which the loss of preferred electrical power supply and failure of automatic reactor shutdown occurred, are presented in Figure 20. The analysis was performed using RELAP5 model, presented in Figure 6.

Due to the loss of preferred electrical power supply all pumps are switched off (see Figure 20 (a)); therefore, the coolant circulation through the fuel channels is terminated. Because of the lost circulation, fuel channels are not cooled sufficiently and for this reason, the temperature of the fuel channel walls starts to increase sharply. As it is seen from Figure 20 (b), already after 40 seconds from the beginning of the accident, the peak fuel channel wall temperature in the high power channels reaches the acceptance criterion of 650 oC. It means that because of the further increase of the temperature in the fuel channels, plastic deformations begin, i. e. because of the influence of internal pressure, the channels can be ballooned and ruptured. During the first seconds of the accident, the main electrical generators and turbines are switched off as well. Steam generated in the core is discharged through the steam discharge valves, but their capacity is not sufficient Therefore, the pressure in the reactor cooling system increases and reaches acceptance criterion 10.4 MPa approximately after 80 seconds from the beginning of the accident (see Figure 20 (c)). Further increase of the pressure can lead to a rupture of pipelines.

Thus, the analysis of the anticipated transients without the shutdown demonstrated that in some cases the consequences can be quite dramatic for the RBMK-1500 reactors. Hence, in 1996 the priority recommendation was formulated as follows: to implement a second diverse shutdown system based on other principles of operation,. The implementation of such system requires much time and financial sources, thus at first it was decided to implement a compensating measure: a temporary shutdown system. This temporary system was called by the Russian abbreviation „DAZ" („Dopolnitelnaja avarijnaja zacita" — „Additional emergency protection"). This system used the same control rods as well as design reactor shutdown system, however, signals for this system control were generated independently in respect of the design reactor shutdown system. The analysis performed to justify the selected set points for reactor scram activation showed that after the implementation of DAZ system, the reactor is shutdown on time and cooled reliably; moreover, the acceptance criteria are not violated even in case of transients when the design reactor shutdown system does not function. Figure 20 presents the behavior of the main parameters of the reactor cooling system in case of the loss of the preferred electrical power supply and simultaneous failure of the design reactor shutdown system [23]. In this case two signals for activation of DAZ system (reactor shutdown) are generated: on the increase of pressure in the drum-separators and on the decrease in the coolant flow rate through the main circulation pumps. In Unit 1 DAZ system was installed in 1999, in Unit 2 in 2000. Later (in 2004) the second diverse shutdown system
was installed in the Ignalina NPP Unit 2. After these modifications the frequency of ATWS at Ignalina NPP became negligible (<10-7/year).

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The GDH blockage for RBMK-1500 also depends on such group of accidents when during normal operation a group of fuel channels is overheated and multiple rupture of FC can occur. It is shown [12] that the coolant flow through the ECCS bypass line is not enough to cool down the fuel channels connected to the blocked GDH. The critical heat flux would appear in some fuel channels and cause failure of fuel claddings and FC walls. In the year 2000 a new reactor scram (emergency shutdown) signal based on coolant flow rate decrease through GDH was implemented at Ignalina NPP. The new signal ensures the timely reactor shutdown so that the dangerous fuel cladding and FC walls temperatures are not reached [12]. Therefore, this accident is moved from the group of severe accidents into the group of accidents without core damage.

The accidents when the loss of natural circulation occurs due to a sharp decrease of pressure in the RCS (due to break of steamlines) are presented in section 5.2.4. Some parts of steamlines are located in the compartments without pressure gauges. Thus, there is no direct signal indicating that the steamline break occured in these compartments (as in the other cases the pressure increase in compartments indicates coolant discharge through the break). It means that signals for the reactor shutdown and ECCS activation will be generated with delay on the basis of secondary parameters (e. g., water level decrease in DSs). On the other hand, a sharp pressure drop in the RCS is a characteristic feature in the case of RBMK steamline break; it destroys the natural circulation of coolant through the core. The flow stagnation in the core together with the late reactor shutdown can cause overheating of group of fuel channels. This was mentioned in the safety analysis report of Ignalina NPP [3] and the review of safety analysis report [24]. In the year 1998 — 1999 a new reactor scram signal based on fast pressure decrease in DS was implemented at Ignalina NPP. This modification allowed avoiding the overheating of a group of fuel channels.

The designed system was tested to confirm the performance and application conditions of a remote distance measurement sensor selected.

The tests were carried out repeatedly to confirm reliability, consistency, accuracy and stability of the measurement.

The reliability test method of the sensors is as follows. First, a sensor was fixed to an anchor of granite comparator stand. Second, the probe of the sensor is positioned so as to be close to the granite comparator stand at suitable heights. Third, the probe was extended so that it touches on the face of stand. This state is the zero point of the sensor. Forth, put gauge blocks of 5 [mm], 10 [mm], 15 [mm], and 18 [mm] on the stand, and measure the reliability of the sensor. Reliability test results were obtained as shown in Fig. 7.

As the size of a gauge block is large, the error of measurement sensor was increased from 0.0037 [mm] to 0.0119 [mm]. However, the reliability test was satisfied because the sensors did not exceed the maximum allowable error (0.0254 [mm]). And the measurement errors under the pressure of 0.8 [bar] and 2 [bar] are similar.

The consistency test method of connection jigs for sensors is as follows. First, a sensor was inserted into a threaded connection jig and fixed firmly. Second, insert a sensor and a connection jig to the hole of the minimized model of the CSB snubber lug, and fix it firmly by threaded connection jig. Third, when sensors have received air pressure of 0.8[bar], 1.4[bar], and 2.0[bar], the sensors measure the distance five times. Fourth, the process measurements are repeated three times and attached again after removing the connection jigs of the remote distance measurement sensors. Consistency test results under 1.4[bar] at each gauge blocks were obtained as shown in Fig. 8.

The measurement errors of sensor 2 (0.0024 [mm]) and 4 (0.0025 [mm]) were large compared with sensor 1 and 3, but consistency test were satisfied because the error did not exceed the maximum allowable error.

The accuracy test method for the zero point adjustment device for the sensors is as follows. First, a zero point adjustment device that was designed and made, binds the right and left of the minimized model of the CSB, and is fixed. Second, run the probes of the sensors installed into the minimized model of the CSB, and remove the zero point adjustment device on the minimized model of the CSB after repeating the distance measurement five times. Third, compare the distance measurements from three repetitions of the process. Accuracy test results were obtained as shown in Fig. 9.

The measurement errors of sensor 2 and 4 were large compared with sensor 1 and 3, but accuracy test were satisfied because it did not exceed the maximum error.

Switching noises and EMI (electromagnetic interference) might occur because of the use of electric lamps and transceivers to build nuclear power plants (Ko & Bae, 2006).

Therefore, the stability tests were done at no disturbance noise and in a switching noise environment, and at EMI environment.

The stability test method of sensors at no disturbance noise is as follows. The air pressure to 1.4[bar] on the remote sensors of the reduced-scale model system was set up and tested five times, repeatedly. Stability test results at remote distance were obtained as shown in Fig. 10.

There were no errors in all sensors except sensor 4. The error was only 0.0003 [mm], but it was in allowable range.

The stability test method of the sensors in a switching noise environment is as follows. First, a 27[W] desk lamp (220V/60Hz) at 10 [cm] from the sensor probe was installed. Second, five times while turning the power on the lamp on and off were measured. During this test, the supplied air pressure was 0.8[bar] to the the sensor probe. The results at switching noise environment were obtained as shown in Fig. 11.

The measurement errors of sensor 2 and 4 were 0.0012 [mm] and 0.0008 [mm], but it were satisfied because the sensors did not exceed the maximum error (0.0254 [mm]).

Another stability test methods of the sensors at EMI environment is as follows. First, a VHF/UHF FM radio transceiver (TM-V7A/KENWOOD) and an antenna 30 [cm] away from the probes were installed. Second, five times each for the cases of 144[MHz/5W], 144[MHz/10W], 439[MHz/5W] and 439[MHz/10W] were repeatedly measured. During this test, the supplied air pressure was 0.8[bar] to the probes of the sensors. The test results at EMI environment were obtained as shown in Fig. 12.

In the environment of 144 [MHz/10W], the measurement error was 0.0003 [mm], but it was in allowable range also.

Fig. 12. Stability test results at EMI environment

As shown in the above test results, the selected sensor can be used in a remote measurement system for the modularization of reactor internals since the sensor errors did not exceed 0.0254 [mm] (1/1000").

In the experiments and results, they were found that the measurement errors of sensor 2 and sensor 4 are bigger than sensor 1 and sensor 3. We judged that these results were occurred by something problem from self-characteristics of sensors.

1.2 Conclusion

From these results, the technology of remote measurement for the modularization of reactor internals may be advanced by design and development of the reduced-scale model system.

Also, the reduced-scale model system designed may be used as a gap measurement training system for a modularization method of reactor internals. And it needs the more suitable sensor under the consideration of the special conditions and environments in reactor internals.

The OSU-MASLWR test facility (Modro et al., 2003; Reyes & King, 2003; Reyes et al., 2007; Galvin, 2007; Mascari et al., 2011a, 2011b, 2011c, 2011d, 2011e), figure 2, is scaled at 1:3 length scale, 1:254.7 volume scale and 1:1 time scale, is constructed entirely of stainless steel and it is designed for full pressure and full temperature prototype operation.

The facility includes the primary and secondary circuit and the containment structures. Two vessels, a High Pressure Containment (HPC) vessel and a Cooling Pool Vessel (CPV) with an heat transfer surface between them to establish the proper heat transfer area, are used to model the containment structures, in which the RPV sits, as well as the cavity within which the containment structure is located. The two middle ADS lines, two high ADS lines and the two ADS sump recirculation lines are modelled separately. In addition to the physical structures that comprise the test facility, there are data acquisition, instrumentation and control systems.

The facility is instrumented for capturing its thermal hydraulic behaviour during steady and transient conditions; in particular thermocouples are used to measure fluid, wall and heater temperature; pressure transducers are used to measure pressure; differential pressure cells are used to measure water level, pressure loss and flow rate; flow Coriolis meters are used to measure mass flow rate; vortex flow meters are used to measure steam mass flow rate, pressure and temperature; power meters are used to measure heater power.

In the previous testing program four tests have been conducted: the OSU-MASLWR-001 — inadvertent actuation of 1 submerged ADS valve-; the OSU-MASLWR-002 — natural circulation at core power up to 210 kW-; the OSU-MASLWR-003A — natural circulation at core power of 210 kW (Continuation of test 002)-; the OSU-MASLWR-003B -inadvertent actuation of 1 high containment ADS valve-.

Since the target of the OSU-MASLWR-001 test was to determine the pressure behavior of the RPV and containment following an inadvertent actuation of one middle ADS valve, it gives a wide number of informations about the primary/containment coupling phenomena characterizing the MASLWR design. Therefore it is the test chosen for this analysis.

The results of the temperature measurement are to be processed quickly in order to get a first estimation of the fatigue state. One important task before the simplified and automated evaluation is the verification of the acquired data. Detection and adjustment of implausible data are parts of this process. In this context, data plausibilization is based on the limits of the measurement range (e. g. 0°C — 400°C) as well as the predefined limiting gradients. Irregularities such as those resulting from the switching of the main coolant pump by electromagnetic pulses are recognized and corrected this way. The original set of data is not modified. All adaptations are reproducibly recorded. These plausibility and quality checks of the measured data have to be done by experienced specialists. In other words, the specialists must be capable of checking the operational events with respect of their plausibility. The result is a preprocessed database for data evaluation and fatigue assessment.

In the very first step of the SFE, the changes of temperatures are subject to a rain-flow cycle counting algorithm (see e. g. [9] and [10]). In this process the temperature ranges at the locations of measurement are identified, counted and classified. The according temperature differences between a subsequent minimum and maximum are inserted into a rain-flow matrix. The temperature difference, the starting value and the temperature change as well as the stratification differences are the parameters of this matrix. An exemplary matrix is shown in Figure 5.

These thermal load cycles are input data for a stress and fatigue assessment of the monitored components based on conservative analytical computation formulae.

The temperature differences are processed to stresses by applying the equation of the completely restrained thermal extension a=E-a-AT (a… Stress, E… Young’s modulus, a… coefficient of thermal extension, AT… temperature difference). The internal pressure induced stresses are added. The stresses calculated in this way are multiplied by stress concentration factors as a function of the component geometry. The well-known stress concentration factors for pipe bends, t-joints or weld seams (see e. g. chapters on piping design in the ASME code [1] or the KTA rules [4]) are applied. Based on these stresses and their frequency of occurrence the partial usage factors are calculated and summed up to the total usage factor. The basic function of this method is a check of fundamental fatigue relevance of the component subjected to the recorded loading. It constitutes a simple qualitative assessment method. In case of calculated CUFs < 5 per cent there is no fatigue relevance of the component. Additionally, SFE allows for a simple qualitative comparison of annual partial usage factors. By means of extrapolation the future fatigue potential can be predicted.

SFE has been successfully applied in many German NPPs for about 20 years. This rough real time fatigue estimation is done after every operational cycle and allows for a direct comparison of thermal loads and an evaluation of the current fatigue usage factor. The result of this SFE provides a qualitative tendency. Although the correlation of the real temperature ranges is fairly simple, it is suitable for a comparison of different real sequences of loads and allows for a qualitative evaluation of the mode of operation and the detection of fatigue critical locations. Furthermore, the investigation of the results allows for the detection of anomalies.