Volume 76, Issue 5

Starting from early to mid 1970s the first generation of channel-type graphite-moderated boiling water reactors socalled RBMK types were designed and went into operation. Nowadays, all designed RBMKs represent three generations of reactors having significant differences with respect to their safety design features. The Ignalina NPP with two RBMK-1500 units being a second generation type were originally the most powerful reactors in the world. The change of important neutron-physical characteristics, being specific safety features for RBMK type reactors during RBMK-1500 core upgrading processes are analyzed until final shutdown and their impact on safety during reactor lifetime is discussed.

Unit 1 and Unit 2 reactors of the Ignalina Nuclear Power Plant were shutdown in December 31, 2004 and December 31, 2009, respectively. Preliminary and Final Decommissioning Plans for the Ignalina Nuclear Power Plant Units 1 and 2 were prepared, the latter of which was approved by Ministry of Economy in 2005. The main points of Final Decommissioning Plan and the main Decommissioning Projects for the Ignalina Nuclear Power Plant are summarized in this paper. Particular attention is drawn to the spent nuclear fuel and radioactive waste management projects. Preliminary analysis showed that the main mass of equipment to be dismantled is concentrated in reactor buildings (∼39%) and turbine halls (∼37%) and consists from metals of different types (∼83%). From radiological point of view, a major part (∼61%) of contaminated equipment can be classified as very low level radioactive waste.

New solid radioactive waste management and interim storage facilities will be constructed for the Ignalina Nuclear Power Plant to support ongoing decommissioning activities, including removal and treatment of operational waste from the existing storage buildings. The paper presents approach and methods that have been used to assess radiological impacts to the general public potentially arising under normal operation and accident conditions and to demonstrate compliance with regulations in force. The assessment of impacts from normal operation includes evaluation of exposure arising from release of airborne radioactive material and from facilities and packages containing radioactive material. In addition, radiological impacts from other nearby operating and planned nuclear facilities are taken into consideration. The assessment of impacts under accident conditions includes evaluation of exposure arising from the selected design and beyond design basis accidents.

The proposed Advanced Heavy Water Reactor (AHWR) employs double containment envelope along with many Engineered Safety Features (ESFs) to mitigate the consequences of Loss-of-Coolant Accidents (LOCA) with safety system failure, during which high enthalpy steam and radioactive fission products will be discharged into the containment. In such conditions, the pressurized containment will be the source of activity release to the environment by way of leakage. It is required to study the effect of ESFs on the source term from the AHWR containment. An analysis was performed to evaluate the release rate from the AHWR containment during a postulated accident with the in-house containment code CONTRAN and the aerosol behavior code NAUA5-M in a coupled way. Modules for simulating the engineered safety features were incorporated in the CONTRAN code and the aerosol transport behaviour was evaluated using NAUA5-M separately. The AHWR containment is divided into three nodal volumes interconnected by junctions. The blow down mass, energy discharge data and activity released into the containment from the reactor core, for a postulated LOCA case of 200% RIH break with failure of shutdown systems (1 & 2), are inputs to the CONTRAN code. Thermodynamic parameters like containment gas temperature, partial pressure of steam, air in the subdivided volumes along with the flow rates through junctions obtained from CONTRAN were supplied to NAUA5-M. An analysis was carried out for a number of cases, postulated based on availability/unavailability of ESFs. Pressure, temperature and activity concentration transients were evaluated, for 72 h, in the subdivided volumes along with the activity released out of the containment through leakages and stack discharges for all the cases. This paper highlights the importance of operation of ESF in reducing the activity release to the environment.

In this study, the time dependent criticality and burn up analyses for a full core model of Pebble Bed Modular Reactors, such as PBMR-400, are performed using weapon grade plutonium mixed with thorium as an alternative to enriched UO 2 fuel. Neutronic calculations are performed with the MCNP and MONTEBURNS 2.0 using ENDF/B-V nuclear data library. 9.6% enriched UO 2 gives an effective multiplication factor of k eff ≈ 1.2395, which can be achieved by 18% weapon grade plutonium +82% ThO 2 fuel mixture. Thorium mixed fuel with 18% weapon grade plutonium fuel would allow reactor operation periods ∼7 years without fuel change with burn up levels of 205000 MW.D/t, for end of life k eff set equal to 1.02.

Selection of first wall material for fusion reactors is very crucial when taking into account of fusion blanket design and operation cost. A realistic way to extend the working period of first wall structure is to use a protective flowing liquid wall between fusion plasma and first wall. HYLIFE-II, one of the important fusion reactor design concepts, uses such a liquid wall. In the current article, the radiation damage on the first wall of HYLIFE-II fusion reactor was investigated for various candidate materials. In the liquid wall of the reactor, a molten salt containing weapon grade (WG) plutonium isotopes was used. The numerical results indicated that a refractory alloy of W-5Re was found to have the lowest damage values. In addition, the use of WG plutonium isotopes did not have a negative effect on the radiation damage characteristics of the investigated structural materials.

In this study, the steady and transient neutronic behaviour of MYRRHA ADS is investigated. For this purpose, a recently proposed analytical benchmark of the diffusion kinetics as 1D slab model of the MYRRHA ADS concept developed in Belgium has been extended to the cylindrical geometry which represents the system more realistically. Analytical calculations are performed using the Customized Solution Method and numerical Laplace inversion techniques such as Fixed-Talbot and Gaver-Wynn-Rho algorithms. Results are compared with the finite element program FLEXPDE® and they are found to be in complete agreement. The necessity of modeling the MYRHHA reactor in cylindrical geometry rather than slab geometry to obtain more realistic benchmark results is demonstrated.

For roughly 25 years, 125 I sources have been used in the treatment of various malignant diseases such as prostate cancer. Three new brachytherapy sources, IR01- 125 I, IR02- 125 I and IR03- 125 I, have been developed and are designed for permanent implant application. The Monte Carlo radiation transport code version MCNP 5 was used to calculate the dosimetry parameters around the sources in accordance with the updated report of the American Association of Physicists in Medicine (AAPM), Task Group No. 43. For each source, the dose rate constant Λ, the radial dose function g L (r), and the anisotropy function F(r, θ), were obtained. The results indicated a dose rate constant of 0.932 ± 0.01, 0.934 ± 0.01 and 0.939 ± 0.01 Gy h −1 U −1 for the IR01- 125 I, IR02- 125 I and IR03- 125 I sources respectively. With the goal of determining an optimal design for a 125 I source, each seed's parameters were compared with other seeds. In this study, the optimal source IR03- 125 I provides the most isotropic dose distribution in water. Finally, the results for optimal source were compared with published results for those of other commercial sources.

89 Zr is an important positron-emitting radionuclide for positron emission tomography (PET) and used in the field of tumor diagnostics, tumor therapy and the investigation of the bio-kinetic. The ALICE-91 and TALYS-1.0 codes were used to calculate excitation function for proton, alpha, deuteron and neutron induced on various targets that lead to the production of 89 Zr radioisotopes using intermediate energy accelerators. Requisite thickness of the targets was obtained by SRIM code. The 89 Zr production yield was evaluated using excitation function and stopping power. For the 89 Y(p, n) 89 Zr reaction, the calculation data were compared with the experimental data. The 89 Y(p, n) 89 Zr process was determined as most interesting one due to radionuclide purity. The ALICE-91 and TALYS-1.0 codes predict a maximum cross-section of about 939.86 mb at 15MeV and 860 mb at 13MeV respectively for this reaction.

Derived from AAPM task group No. 60/149 protocol, applicable in treatment planning In this study, the two-dimensional dose distributions in water for a 32 P intravascular brachytherapy stent have been calculated. The pure beta emitter source 32 P which has been coated on Palmaz-Schatz stent is discussed. The dosimetric parameters required by the AAPM TG-60/149 formalism are discussed and calculated. Version 5 of the (MCNP) Monte Carlo radiation transport code was used to calculate the dosimetry parameters around the source. The Monte Carlo calculated dose rate at the reference point is found to be 2.8 Gy/μCi. Also in this study, the geometry function, G(r,θ), radial dose function, g(r), and the anisotropy function, F(r,θ), have been calculated at distances from 1.8 to 9 mm. The results of these calculations have been compared with other published calculated and measured values for an actual same source. High dose variants were visible near the 32 P stent surface, but these values decreased with depth in water rapidly. There is an acceptable agreement between the calculated data in this study and other published data for the same source, which validate our simulations method.

Excitation functions for protons induced on 85/nat Rb targets that lead to the production of 85 Sr radioisotopes using TALYS 1.0 and ALICE/ASH codes were calculated. Rubidium chloride deposition on copper substrate was carried out via sedimentation method in order to produce Strontium-85. 520 mg RbCl, 208 mg ethyl cellulose (EC) and 4 mL acetone were used to prepare a layer of enriched rubidium chloride of 11.69 cm 2 area and 62.2 mg · cm −2 thickness. The deposited target was irradiated at 20 A current and 15 MeV proton beam for 10 h and no degradation was observed. The 85 Sr production yield was 1.4 MBq/mA · h. The target material was dissolved by 50 mL acetone. Dissolved Strontium-85 was separated from rubidium by Chelex 100 ion exchanger.

Customer requirements to probabilistic safety targets are usually stronger than existing Regulatory or IAEA ones. It appears that industry takes the lead over regulation in this case and forces the designer to find and implement appropriate means to enhance safety, which sometimes have no reference to practical experience. On the other hand, regulatory documents and the existing PSA methodology are mainly oriented to operating plants. This creates problems when developing a PSA as well as performing regulatory reviews. The scope of the PSA may be different depending on a design stage such as the development conceptual, basic or detailed design. In addition, the base case PSA is usually performed for NPP in design. However, a customer may require additional PSA applications to consider, for instance, risk monitoring. In this case the scope of the PSA should be extended to implement special attributes of the application needed that often requires specific information not available at the design stage. Lack of design information affecting PSA development may be associated with incompleteness of the design that is typical for interim design stages and communication problems caused by the involvement of many different companies in the deign activity. To deal with this issue bounding technologies and the iterative PSA development are used. However this sometimes contradicts to the “best estimate” approach recommended by regulatory guides. PSA development for advanced NPPs has raised some issues originated from unknown new components, processes and technologies incorporated into the design of an advanced plant. The paper addresses some issues resolved while carrying out PSAs for advanced NPPs. Some PSA results for new advanced VVER plants under construction and the first lessons learnt from the Fukushima accident are also discussed.