1. In order to set up the remote measurement system inside the CSB assembly, three channel boxes were mounted in the CSB assembly. As shown in Fig. 15, the channel boxes were opened and 72 digital probes were taken out of the CSB assembly through spaces between the CSB and the LSS.

2. As shown in Fig. 24, all 72 digital probes were installed the holes of the CSB snubber lugs after the digital probes were assembled with the threaded connection jigs. For this step, the digital probes were necessarily placed at 1 mm or more from the measured holes of the CSB snubber lug. Bolt tightening was done using a hexagonal wrench. The bolts numbered 1, 2, 3, and 4 were clockwise and the tightening order was 1, 3, 2, and 4. All remaining sensors were installed via the methods described above.

3. The electric power cords, air hoses, and signal cables between the channel boxes were connected and taken out of the RV. The air hoses were connected with an air compressor.

4. The signal cables were linked with a USB-orbit module and an RS-485 converter. The USB ports of the USB-orbit module were linked with a USB hub. The USB hub and USB ports of the RS-485 converter were connected to the USB ports of a remote measurement computer. Table 5 shows the remote measurement system guideline.

5. Electric power supplies were checked in order with the remote measurement system first, followed by the air compressor, and then the remote measurement computer to verify whether or not the connections worked exactly according to the operating guide.

6. The air pressure was set to 0.8 — 1 bar using a software program on the remote measurement computer. At this stage, the digital probes had to be checked using a software program to verify whether or not they operated normally. Fig. 14 shows the configuration of the remote measurement system.

A nodalization, representing an actual system (integral test facility or nuclear power plant), can be considered qualified when it has a geometrical fidelity with the involved system, it reproduces the measured nominal steady state conditions of the system, and it shows a satisfactory behavior in time-dependent conditions.

The OSU-MASLWR nodalization qualification process (Bonuccelli et al., 1993) is still in progress, because the facility experimental characterization will be conducted in the framework of the current IAEA ICSP. In particular several important facility operational characteristics, like pressure drop along the primary loop, at different primary side mass flow rates, and heat losses, at different primary side temperatures, determined to be of importance during the planned ICSP experiments, will be evaluated and distributed to the ICSP participants. Besides, some nodalization models, here presented, are still preliminary because some geometrical data and the complete instrument characterization and location will be delivered in the ICSP framework as well. Therefore 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 primary/ containment coupling phenomena typical of the MASLWR prototypical design (Mascari et al., 2009b; Mascari et al., 2011c).

The phenomenon of the emission and absorption of a y-ray photon without energy losses due to recoil of the nucleus and without thermal broadening is known as the Mossbauer

effect. Its unique feature is in the production of monochromatic electromagnetic radiation with a very narrowly defined energy spectrum that allows resolving minute energy differences [2,3].

Mossbauer spectroscopy (MS) is a powerful analytical technique because of its specificity for one single element and because of its extremely high sensitivity to changes in the atomic configuration in the near vicinity of the probe isotopes (in this case 57Fe). MS measures hyperfine interactions and these provide valuable and often unique information about the magnetic and electronic state of the iron samples, their chemical bonding to co-ordinating ligands, the local crystal symmetry at the iron sites, structural defects, lattice-dynamical properties, elastic stresses, etc. [1,4]. Hyperfine interactions include the electric monopole interaction, i. e., the isomer shift, the electric quadrupole interaction, i. e., the quadrupole splitting, and the magnetic dipole (or nuclear Zeeman) interaction, i. e., hyperfine magnetic splitting. These interactions often enable us detailed insight into the structural and magnetic environment of the Mossbauer isotope. Indeed, more than four decades after its discovery (1958), Mossbauer spectroscopy still continues to develop as a sophistical scientific technique and it is often the most effective way of characterizing the range of structures, phases, and metastable states.

In general, a Mossbauer spectrum shows different components if the probe atoms are located at lattice positions, which are chemically or crystalographically unequivalent. From the parameters that characterise a particular Mossbauer sub-spectrum it can, for instance, be established whether the corresponding probe atoms reside in sites which are not affected by structural lattice defects, or whether they are located at defect-correlated positions. Each compound or phase, which contains iron, has typical parameters of its Mossabuer spectrum. It means, the method is suitable for quantitative as well as qualitative analysis. Mossbauer spectroscopy is non-destructive and requires relative small quantities of samples (~100 mg) [5-8].

Application of Mossbauer spectroscopy for precise analysis of phase composition of corrosion products was performed from selected areas of primary and secondary circuit and SG. Interpretation of measured results, having in vision the long-term operation and nuclear safety, is not easy, nor straightforward. Thanks to our more than 25 years of experiences in this area, there exists already a base for the relevant evaluation of results. Optimisation of operating chemical regimes as well as regimes at decontamination and passivation seems to be an excellent output.

Process Modeling: The main problem in setting up a signal flow diagram for a level controlled system in a SG can be found in the inhomogeneous contents of the SG.

The filling consists of water at boiling temperature, pervaded by steam bubbles.

Since the volume fraction of the steam bubbles is quite considerable, the mean specific weight of the contents is very strongly dependent on the proportion of steam.

This, of course, means that the steam content also strongly influences the level in the SG. The steam content itself depends, in turn, on the load factor, on the changes in feed-water flow, and on feed-water temperature.

The presence of steam below the liquid level in the SG causes the shrink-and-swell phenomenon that in spite of an increased supply of water, the water level initially falls. Figure 2 shows responses of the water level to steps in feed-water and steam flow-rates at different operating power levels (Irving et al., 1980).

Particularly it is difficult to control automatically a steam generator water level during transient period or at low power less than 15% of full power because of its dynamic characteristics.

The inverse response behavior of the water level is most severe at low power (5%).

The changing process dynamics and the inverse response behavior significantly complicate the design of an effective water level control system.

A solution to this problem is to design local linear controllers at different points in the operating regime and then applies gain-scheduling techniques to schedule these controllers to obtain a globally applicable controller.

Consider a step in feed-water flow rate at 5% operating power. For this system, a fuzzy convolution model consisting of four fuzzy implications is developed as follows:

For j=1 to 4:

R’ : if VDd (n) is A

200

then j (n +1) = y1od () +Xh_Ddu(n +1 -i)

i=1

(a) (b)

Figure 3. shows the response of water level at 5% operating power to a step in feed-water flow — rate. In Figure 4 the system is decomposed into 4 subsystems: yD, , Уг>1, yD t, yD, ■

Figure 5 shows the impulse response coefficients for yD (, yD t, yB yD ( subsystems and Figure 6 shows the definition of fuzzy sets A1, A2, A3 and A4. Consider a step in steam flow rate at 5% operating power.

For this system, a fuzzy convolution model consisting of four fuzzy implications is developed as follows:

For j=1 to 4:

Rj: if Vd0 (n)isAl

200

then yD (n +1) = yD0 (n) + Xh — Do’u (n +1 -1)

0 i=1

Figure 7 shows the response of water level at 5% operating power to a step in steam flow — rate.

In Figure 8 the system is decomposed into 4 subsystems: yD0 , yD0, yD0 , yD0 •

Figure 9 shows the impulse response coefficients for yD0, yD0, yD0, yD0 subsystems, Figure 10 shows the definition of fuzzy sets A1, A2, A3 and A4.

Controller Design: The goal is to study the use of the feed-water flow-rate as a manipulated variable to maintain the SG water level within allowable limits, in the face of the changing steam demand resulting from a change in the electrical power demand.

The simulations are organized around two different power transients:

• a step-up in power from 5% to 10% (Figure 11);

• a ramp-up in power from 5% to 10% (Figure 12)

The model horizon is T=200. Increasing Ny results in a more conservative control action that has a stabilizing effect but also increases the computational effort.

The computational effort increases as Nu is increased. A small value of Nu leads to a robust controller.

For both power transients the controller responses are very satisfactory and not very sensitive to changes in tuning parameters.

We can see that the performance is not strongly affected by the presence of the feed-water inverse response, only a slight oscillation is visible in the water level response.

Fig. 11. Water level response to a step power increase from 5% to 10% (Nu=2, Ny=3, W1=1)

(b) (c)

(d)

3.3 Modeling two-phase flow

Most of the early experimental research in this field relied on sectional models of tube arrays subjected to single-phase fluids such as air or water, using relatively inexpensive flow loops and wind tunnels. The cheapest and simplest approach to model two-phase flow is by mixing air and water at atmospheric pressure. However, air-water flows have a much different density ratio between phases than steam-water flow and this will affect the difference in the flow velocity between the phases. The liquid surface tension, which controls the bubble size, is also not accurately modeled in air-water mixtures. Table 8 gives the comparison of liquid and gas phase of refrigerants R-11, R-22 and air-water mixtures at representative laboratory conditions with actual steam-water mixture properties at typical power plant conditions (Feentra et al., 2000). This comparison reveals that the refrigerants approximate the liquid surface tension and liquid dynamic viscosity of steam-water mixtures more accurately than air-water mixtures.

Typical nuclear steam generators such as those used in the CANDU design utilize more than 3000 tubes, 13mm in diameter, formed into an inverted U-shape. In the outer U-bend region, these tubes are subject to two-phase cross-flow of steam-water which is estimated to be of 20% quality. It is highly impractical and costly to perform flow — induced vibration experiments on a full-scale prototype of such a device so that small-scale sectional modeling is most often adopted. R-11 simulates the density ratio, viscosity ratio and surface tension of actual steam — water mixtures better than air-water mixtures and it also allows for localized phase change which air-water mixture does not permit. While more costly and difficult to use than air-water mixture, R-11 is a much cheaper fluid to model than steam-water because it requires 8% of the energy compared with water to evaporate the liquid and operating pressure is much lower, thereby reducing the size and cost of the flow loop (Feentra et al., 2000).

The analytical study should provide a systematic comparative assessment of the consequences (costs, benefits, impacts and risks) of alternative energy options (technologies). For decision-making purposes, these results need to be evaluated and presented in a coherent way. The evaluation and presentation of the results should focus on pointing out the main findings and conclusions that could support decision taking.

In order to assist decision makers effectively, analysts should present their results in a transparent manner (no "black boxes"), focusing on the verifiable results and their interpretations. In particular: input data and assumptions should be specified clearly and the boundaries and limits of the study should be indicated; comparison of alternatives should be based upon indicators that have been estimated quantitatively and qualitatively.

In general, the presentation of the results has to be adapted to the target audience of the study. The primary audience will be decision makers. However, in most cases, the study will also be disseminated to, and used by, interested and affected parties, e. g. local communities or NGOs. In both cases, the audience has not the same experience and knowledge on technical and economic issues as do the analysts. Therefore, results should be presented clearly and concisely, pointing out the main findings and outcomes.

The sequential analyses determine in addition to 59Ni and 63Ni others DTM’s present in the nuclear waste including alpha emitters. Therefore, alpha spectrometry is one complementary technique for the nuclear waste characterization either ILW or LLW.

In this technique to achieve results with good quality, the sample must be converted into a chemically isolated, thin layered and uniform source. The preparation of an alpha sample contains three basic steps: preliminary treatment, chemical separation and source preparation

Alpha-emitting radioisotopes spontaneously produce alpha particles at characteristic energies usually between about 4 and 6 MeV. Alpha particles (or 4He nuclei) are heavy charged, large and slow particles and loses some of its energy each time it produces an ion (its positive charge pulls electrons away from atoms in its path), finally acquiring two electrons from an atom at the end of its path to become a complete helium atom. These attenuation characteristics, which manifest themselves both within the sample and with any materials between the sample and the active detector volume, cause a characteristic tailing in the alpha peak. When tailing occurs (it is also called "spill down"), the accuracy with which the peak areas can be determined is compromised because the peaks tend to have an asymmetric shape rather than the Gaussian shape.

The alpha particle energies of many isotopes differ by as little as 10 to 20 keV (Canberra, n. d.). The relatively small difference in alpha particle energy between some alpha emitters makes it difficult to spectrometrically separate the peaks once this is near the resolution of the silicon detectors used in alpha spectrometers. If two of these alpha particle energies are so close, they cannot be spectrometrically separated and if they are chemically the same, they cannot be chemically separated and analyzed.

Resolution is the ability of the spectrometry system to differentiate between two different alpha particles and its quantitative measure is the FWHM. Besides, a FWHM of about 15 keV can be achieved with electroplated sources because they have very little mass to slow down the alpha particles. For this reason it is essential that a thin source to be prepared in alpha spectrometry.

The consequences of LOCA in Zone 2 are similar to the consequences of LOCA in Zone 1. In both cases the break location is upstream the reactor core. Moreover, the phenomena in both cases are similar. In case of GDH break the coolant supply is terminated through 39 — 43 fuel channels connected to the distribution header of this group. In case of GDH guillotine break downstream check valve, the coolant in FCs connected to the affected GDH starts to flow in the opposite direction from DSs (Figure 16). Loss of the coolant from DS is compensated by

ECCS water. Short-term fuel cladding and fuel channel wall temperatures increase is observed at the beginning of the accident due to the coolant flow direction change as in the case of LOCA in Zone 1. During partial breaks of GDH pipe, or in case of guillotine break of one GDH with the failure of check valve in the adjacent GDH (Figure 17), the coolant flow rate stagnation is possible in FCs connected to the affected GDH in this zone. In case of lower water piping break, coolant supply is terminated only into one FC.

There are no pipelines with a diameter bigger than 300 mm, thus the large LOCAs in Zone 2 were not analyzed. For the medium LOCA in Zone 2, the following was considered: GDH guillotine break, GDH partial break resulting into stagnation of coolant flow rate and GDH guillotine break at failure to close the check valve in the adjacent GDH: [2]

will remain intact. The conditions for the reactor long-term cooling remain similar to GDH break: the operation of two ECCS pumps is necessary.

• In the case of GDH break with a failure to close the check valve in the adjacent GDH, ECCS water supply worsens the cooling conditions of the channels connected to the GDH with failed to close check valve. It occurs that ECCS water interferes with the reverse coolant flow rate through these FCs. Stagnation of coolant flow rate is formed in these FCs. However, ECCS water supply helps to fill DSs and to ensure cooling of channels in the intact RCS loop and the channels connected to the 18 GDHs of the affected RCS loop. The analysis of GDH break with failure to close the check valve in the adjacent GDH is carried out at the operation of 1 — 4 ECCS pumps. The results of the analysis showed that for the reliable cooling of FC, connected to other 18 GDHs of the affected RCS loop, it is necessary to have not less than two operating ECCS pumps in long-term cooling subsystem. The channels connected to the GDH with failed to close check valve will be cooled because of radial heat transfer between the adjacent graphite blocks.

For the small LOCA in Zone 2, guillotine and partial breaks of lower water pipe were considered:

• The results of the analysis of guillotine break of the lower water pipe showed that the reactor core is reliably cooled during the first minutes after the accident in this case. One ECCS pump is enough for the reactor long-term cooling. The pump should be started during the first hour after the beginning of the accident.

• The performed analysis demonstrated that in the worst case, at partial break of the lower water pipe, the signal on the reactor shutdown on pressure increase in ALS compartments can not be generated. If the partial break causes stagnation of the coolant flow rate through the affected FC, it will lead to the heat up and break of this channel. The peak temperature of fuel in the affected channel will not reach the temperature of melting, i. e. 2800 °С. After the break of the channel pipe, pressure in the reactor cavity increases, which results in the formation of a signal on the reactor shutdown. After the fuel channel wall break the conditions of flow stagnation will be destroyed and the broken parts of fuel channel and fragments of fuel assemblies below and above the break will be cooled by coolant flow from the top and bottom. The remaining intact fuel channels will also be reliably cooled. It is necessary to note that for RBMK type reactors the rupture of a single FC is a design basis accident. Such accident would correspond to a small breach in the reactor vessel of BWR. Steam-gas mixture from RC will be discharged through the reactor cavity venting system to the left tower of ALS. The steam will be condensed, fission products will be scrubbed in the condensing pool, but will not be discharged to the environment. Thus, the damaged fuel assembly will be contained in the reactor cavity.

Generally, the RVI comprise three components: the core support barrel (CSB), the lower support structure (LSS)/core shroud (CS), and the upper guide structure (UGS). The existing method of assembly is very complicated and requires approximately 8 — 10 months to complete (Korea Electric Power Research Institute, 1997) (Korea Hydro & Nuclear Power Co., Ltd., 2002). The installation of the reactor vessel (RV) is a critical process during the construction period.

This part describes the RVI installation method using the RVI modularization which can shorten the construction period by a minimum of two months compared to the existing method. In order to modularize the RVI, gaps between the CSB snubber lug and RV core- stabilizing lug must be measured using a remote method at outside the RV. Therefore, this part includes explanation on a measuring system to measure gaps between the RV and the CSB remotely with the aim of RVI modularization. The remote measurement system was developed for use at actual construction sites of nuclear power plants using a measurement sensor, a threaded connection jig, and a zero-point adjustment device. With these, a reduced-scale model system was validated. With the remote measurement system, experiments and analyses were performed using mockups for both the RV and the CSB to

verify the applicability of the described system in a construction project. From the data acquired by the remote measurement system, shims were separately made and adjusted.

After installing the shims on RV core-stabilizing lugs, the gaps satisfied requirements within the permissible range of 0.381 — 0.508 mm. The reliability and applicability of the remote measurement method were evaluated and it was concluded that the remote measurement system enables RVI modularization with a significantly reduced construction period.

Fig. 1 shows an existing nuclear reactor installation method and the developed modularization method by remote gap measurement.

RVI is classified on a large scale into three categories: CSB, LSS/CS and UGS. When a nuclear power plant is built, the materials are delivered and assembled according to which category they fall under. It is, however, possible to modularize the installation of the CSB and LSS/CS (Ko et al., 2009) ( ABB-CE, 1995).

Gaps between CSB snubber lug and RV core-stabilizing lug can be measured at outside the RV by a remote method in order to modularize the RVI. If the CSB module (CSB and LSS/CS) is installed into RV, access to measuring the gaps is cut off by the LSS/CS. Therefore, the gaps must be measured remotely at outside the RV.

Fig. 2 shows a picture of the second step in Fig. 1(a)

(b) Proposed modularization method

The hand-measurement of Fig. 2 takes a lot of measurement times and it occurs measurement errors by measurers. Also, a measurement space is small and narrow, and environment to measure gaps between RV and CSB is uncomfortable.

Fig. 3 shows timescales of existing installation method and developed method.

As shown in Fig. 3, developed modularization method by gap measurement remotely can reduce the critical path of the construction period by approximately 8-12 weeks. If the construction period is reduced, a construction budget expected to be saved the minimum $ 176 million in Korea.

Fig. 4 shows a section of the RV core-stabilizing lug and the CSB snubber lug that shall be measured remotely for the modularization of the RVI.

In the beginning tests of control rod drive mechanism (CRDM), two phase flow would come into being because there is air dissolved in the liquid and hidden in the groove of driving rod. The air would be decomposed from the water and be extruded out of the groove when there is disturbance. For example, electrifying of coil component, up-down movement of driving rod, swing in and out of gripper component would give rise to disturbance and make the bubbles appear in the liquid. There is small amount of air dissolved or hidden in the water and two — phase flow would not occur in the test after CRDM moves several days.

In the tests, there is no bubble observed and the air would not be decomposed from the water when there is no disturbance, which is shown in Fig. 21(a). There are three coil components in CRDM, which are lifting coil, moving coil and stationary coil. And bubbles begin to come into being and there appear several separate bubbles when the coil component is electrifying and the gripper swings into the driving rod, which is shown in

In the tests of rod dropping, the power of three coils is cut off, the grippers swing out of the driving rod, and the driving rod free falls in the rod travelling house. In this process, the disturbance is transitory and bubbles appear in the liquid, of which the amount is less than that of rod lifting. When the bubbles go up to the bottom of the test section and there is no disturbance again, no bubbles will generate in the liquid, as shown in Fig. 22.

In the cold tests of rod lifting, rod inserting and rod dropping, only bubbly flow comes into being due to that the amount of gas dissolved in the liquid is small. After CRDM moves several days, the gas dissolved are all driven out of the liquid and there will not appear two — phase flow.

Fig. 22. Rod dropping tests

3. Conclusion

On the basis of the experimental results, the conclusions are obtained,

(1) It is found that there are several main flow patterns, bubbly flow, bubbly-churn flow, churn flow and annular flow in the tube-bundle channel. And there are great differences between flow patterns and their transitions in a tube-bundle channel and that in a circular tube.

(2) Experiments show that there may be two different flow patterns in the same cross­section of the tube-bundle channel. And the flow pattern transitions exhibit unsynchronized in different sub-channels. This unsynchronized phenomenon is caused by the different geometric dimensions, different heat flux and different quantity of discrete bubbles generated in different sub-channels.

(3) The flow pattern map is drawn on the basis of experiments. Comparisons are conducted between flow pattern transition in a tube-bundle channel and that gained by Hewitt & Roberts. The results show that the regions of bubbly flow and churn flow in a tube-bundle channel are larger than that in a circular tube. In addition, the flow pattern transforming to annular flow is earlier in a tube-bundle channel than that in a circular tube.

(4) Three main tests are carried out to measure pressure drop of the containment sump strainers. The photographs are taken and the vortex is not observed in these tests.

(5) In the tests of control rod drive mechanism, two-phase flow is observed. The results of upwards movement and drop movement of the driving rod are compared and analyzed.

4. Acknowledgments

The authors would like to acknowledge the financial support from the National Science and Technology Program of China (Grant No. 2011BAA06B00). The authors would also like to thank Messrs. Yu Jiang., Zhou Jianming, Wu Wei, Bai Bing, Lu Zhaohui for the helpful discussions.

5. References

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Hewitt, G. F. & Roberts, D. N. (1969). Studies of two-phase flow patterns by simultaneous X-ray and flash photography, Atomic Energy Research Establishment, ISSN: 0029-5450, Harwell, England

Lu, G. Y.; Ren, J. S.; Zhang, C. G. & Lin, P. (2011). Debris transport calculation of nuclear power plant containment, 19th International Conference On Nuclear Engineering, ISBN: 978-0-7918-4351-2, Osaka, Japan, October 24 — 25, 2011 Lu, G. Y.; Ren, J. S.; Zhang, C. G. & Lin, P. (2011). Investigation on pressure drop characteristics and disposal optimization of conflux channels of containment sump strainers, 19th International Conference On Nuclear Engineering, ISBN: 978-0-7918­4351-2, Osaka, Japan, October 24 — 25, 2011

Lu, G. Y.; Ren, J. S.; Zhang, C. G. & Xiang, W. Y. (2011). Application and development of containment sump strainers in PWR power stations in China, 19th International Conference On Nuclear Engineering, ISBN: 978-0-7918-4351-2, Osaka, Japan, October 24 — 25, 2011

Ma Weimin. (1992). Experimental investigations on two-phase flow in heat exchangers. Xi’an Jiaotong University, ISSN: 1671-8267

Petigrew, M. J. & Taylor C. E. (1994).Two-phase flow-induced vibration. Journal of Pressure Vessel Technology, Vol.166, pp. 233-253, ISSN: 0094-9930 Sadatomi, M. & Kawahara, A. (2004). Flow characteristics in hydraulically equilibrium two — phase flows in a vertical 2×3 rod bundle channel. International Journal of Multiphase Flow, Vol.30, pp.1093-1119, ISSN: 0301-9322