Band 78, Heft 6

Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s–early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative study of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton® GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or readymade supply. R&D necessities for inflatable seals and their development across the world are given closer look in Part II of the review in continuation of Part I.

Ignalina NPP was operating two RBMK-1500 reactors which are under decommissioning now. In this paper, analysis on radwaste management alternatives during the dismantling of systems with low level contamination and different types of components in buildings 117/1 and V1 are presented. After situation analysis and collection of the primary information related to components' physical and radiological characteristics, location and other data, two alternatives for radwaste management during the dismantling were formulated and evaluated: the first one (A1) when the decontamination of the dismantled components is performed (if it is reasonable), and the second one (A2) when no decontamination of the dismantled components is performed and after the dismantling, the components are routed to appropriate waste storage or disposal sites. To select the preferable alternative, MCDA method – AHP (Analytic Hierarchy Process) is applied. Hierarchical lists of decision criteria, necessary for assessment of alternatives performance, are formulated. Quantitative decision criteria values for these alternatives are calculated using software DECRAD, which was developed by Lithuanian Energy Institute Nuclear Engineering Laboratory. Qualitative decision criteria are evaluated using expert judgment. Analysis results show that alternative A1 has a preference against alternative A2.

Copper and cobalt radioisotopes have been used for both diagnostic and therapeutic purposes in the field of nuclear medicine. There is considerable interest in 60 ,61,64,67 Cu and 55,57 Co radioisotopes for some applications like PET, SPECT imaging and targeted radiotherapy of tumors. In present study, the production of 60,61,64,67 Cu and 55,57 Co via 64 Zn(p,α) 61 Cu, 67 Zn(p,α) 64 Cu, 70 Zn(p,α) 67 Cu, 58 Ni(p,α) 55 Co, 60 Ni(p,α) 57 Co, 64 Zn(p,nα) 6 °Cu and 68 Zn(p,nα) 64 Cu reactions have been investigated in the range of 5–40 MeV proton incident energy. The new version of the code ALICE-2011 is used to determine the excitation functions of these nuclear reactions. The calculated excitation functions were compared with experimental results and semi-empirical cross section values of a previously developed model.

For analyzing the failure probability of the system, a model of the system was developed in best estimate T–H code RELAP5/Mod3.3. The parameters affecting the system performance were identified, their range and distribution was determined. Feasibility of various methods for calculating functional reliability was analyzed. The response surface method was selected. A response surface was constructed for the entire range of parameters for the model developed using Monte Carlo method. A large number of samples of the input parameters was generated, taking into account their individual distributions. They were classified as success/failure based on their position with respect to the failure surface. The ratio of the failed samples to the successful samples gives us the failure probability. An attempt was also made to integrate the functional failure probability with the component failure probability of the system to get an overall failure probability of the system.

Due to the reduction of accessible uranium resources as well as waste proliferation issues, researchers are looking for more suitable approaches, such as replacement of uranium as breeding fuels. Among the practical fuel matrixes, the thorium fuel matrix is favored for its naturally abundant and minor actinide proliferation resistance. Monte Carlo computational methods are widely used to successfully simulate neutronic behavior of nuclear reactors. Calculation of some neutronic and dynamic parameters of a 37-assembly simulated research reactor consisting of thorium oxide fuel and 1 minor actinide pin have been carried out in the present work using the MCNPX 2.6 code.

The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory.

The application of numerical codes is well established in the regulatory practice for nuclear power plants. Therefore, requirements for computer-aided analyses can be found in different places of the nuclear technical rules and standards. These requirements also apply for CFD analyses which reached more and more the spot light of research activities and component design, because computing capacities increased significantly in the past years. Hence, the actual status of CFD as a possible design and analysis tool in the supervising process of nuclear power plants will be discussed in this paper. The focus is set on the prerequisites which have to be fulfilled by codes and in particular by the user who wants to use CFD methods. The prerequisites will be derived directly from the existing German technical rules and standards. Finally, two examples for typical CFD applications for nuclear safety analyses are given.

In this study, the installation and measurement capacity of the neutron irradiation system are investigated. First of all an irradiation geometry enabling optimum irradiation was designed for three 241 Am–Be sources each of it having 592 GBq activity. Neutron irradiation system was installed after design and optimization of the system including the design of appropriate moderator and shielding were completed. Radiation safety standards of the Neutron Research Laboratory fulfilling the requirements of national regulation were achieved with unique configuration of the shielding materials. In this study the results of qualitative and quantitative detection limits obtained for Na, Al, Cl, K, Ti, V, Mn, Fe, Co, Cu, Zn, Ga, Ge, As, Se, Br, Rb, Sr, Y, Mo, Ru, Ag, Cd, In, Sb, I, Cs, Ba, La, Ce, Pr, Nd, Sm, Eu, Gd, Dy, Ho, Yb, Lu, Hf, W, Pt, Au, Th and U elements by using the neutron irradiation cell comprising 3 × 592 GBq 241 Am–Be isotopic neutron source are presented and discussed.

The reflected critical slab problem is investigated in one-speed neutron transport theory for reflecting boundary conditions. The neutron angular flux is expanded in a series of first kind Chebyshev polynomials and isotropic, backward and forward scattering kernel is used as the scattering function. The critical sizes for one-speed neutrons in a uniform finite slab surrounded by identical reflectors from both sides are calculated for various values of the collision parameter, backward and forward scattering and reflection coefficient. Numerical values calculated for the critical slab are presented in the tables and are found to be in good agreement with literature values.