Nuclear Science and Technology, Vol.5, No (2015), pp 01-06 Human Performance in the Nuclear Industry Steven M Koncz Human Performance Engineering Pty Ltd, PO Box 5036, Falcon, Western Australia Email: skoncz@hpeng.com.au (Received 10 September 2015, accepted 25 October 2015) Abstract: Management of employees human performance in the Nuclear Industry is endemic to their safety when working In the United Kingdom it has been a key focus since 2003 Employees were made aware through a detailed program of workshops, of the error prevention methods and how to apply them The use of effective incident barriers became embedded in the safety culture The methodology implemented was personal ownership, to enable self assessment of behaviors, attitudes and beliefs When put in place, there are many specific barriers, which can reduce the chances of an error occurring They come under the headings of organisational, procedural and physical barriers All of these were used in some way and continue to be reinforced on a daily basis Specific barriers are applied in specific situations However, some general ones are also effective In common use are the Take or Take Minutes, point of work risk assessments Applying the human performance barrier Independent Verification (I.V.) would result in 'Take and I.V.' This would independently double check the risk assessment New ways of thinking are required to continuously improve and evolve Results of the error reduction process included; reduced workload, increased plant reliability, efficiencies and productivity Keywords: Error, Human, Performance, Work, Prevention, Nuclear, Barriers, Safety, Process, Behaviour I INTRODUCTION This paper describes the history of human performance error prevention, as used in the nuclear industry How error prevention tools are used and how we could improve on the ways in which they are employed on a daily basis In the 13 years of use at nuclear facilities, it is suggested the error prevention tools have the error prevention tools applied to themselves and review their application to promote continuous improvement 'Complacency' is recognised as one of the error enablers Being comfortable with the way in which the human performance error prevention methodologies are used, is itself an error precursor If we think we have got it right and don't need to change or improve, then we are not applying the tools correctly The 'Norms' is another recognised error enabler, "it's always been done like that," is a reply when asked why a particular action lead to an event of some kind If we fall into the same trap and don't review how we employ the prevention methodologies, we again are not applying the tools correctly II HISTORY OF HUMAN PERFORMANCE IN THE NUCLEAR INDUSTRY The nuclear event on April 26th 1986 at the Chernobyl-4 plant in the then Soviet Union, led to changes in the approach to process safety in nuclear plants the world over The World Association of Nuclear Operators (WANO) was formed on 15th May 1989, under a banner of international cooperation Through open exchange of ©2015 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute HUMAN PERFORMANCE IN THE NUCLEAR INDUSTRY operating experience, all members could then work together to achieve the highest possible standards of nuclear safety were down to the individuals and 70% due to the organisations failing to prevent the errors This is shown in Fig [4] The Institute of Nuclear Power Operations (INPO), founded in December 1979, established a Special Review Committee on Human Performance in late 1993 This committee, along with several working groups, was asked to identify actions to bring about continued improvement in human performance within the commercial nuclear power industry [1] It was this document, which was adopted and reviewed by WANO to form the basis, in 2002, for improving human performance [2] Fig Contribution of human error to the occurrence of events [4] IV WHY CONCENTRATE ON HUMAN PERFORMANCE? III HUMAN PERFORMANCE IMPROVEMENT Human beings are fallible, they make mistakes, and even with the best intentions something can invariably go wrong There is now good evidence through human performance improvement to demonstrate the benefits to safety, production and output “People know the right thing to for any situation in three ways.” First, instinct triggers automatic responses This is a fixed reaction ’hard wired’ in the human mind that elicits a special response, such as the dilation of the eyes as one walks into bright sunlight No learning is required Second, a suitable response is determined by learning either by education, by trial and error, or from others' experiences Examples include reading a book on finances, learning to ride a bicycle, reading operating experience reports, or learning the expectations of a new employer or work group Finally, thinking is a process of building idea upon idea to make sense of a situation Thinking gathers data to generate cues that may help a person recognize a familiar pattern about what to Thinking generates new ideas coupled with new knowledge leads to better understanding [5] In the UK over a 2-year period, the performance of key performance indicators (KPI's) were ahead of WANO “Best in Class” targets for 2004/05 This was attributed to the business improvements at that time Implementing and reinforcing the Human Performance error prevention process had a bearing on these results, Non-outage defects backlog reduced by 55%, Accident frequency rate reduced by 40%, Unplanned automatic trip rate reduced by 30%, Work schedule adherence was 28% better [3] Human error contributes to around 80% of nuclear events in the industry, the remaining 20% attributable to equipment / plant failures This not only has a bearing on the performance of the facilities themselves, but the overall public perception of the nuclear industry Of the identified human errors, 30% of the mistakes The skills, knowledge and attitudes of individuals take time to change It is for this STEVEN M KONCZ Fig Significant Events at U.S Nuclear Plants: Annual Industry Average, Fiscal Year 1992-2006 [6] reason that effective barriers must be put in place Managers implement and strengthen defenses, they reinforce error-prevention techniques and maintain the standards and expectations for staff Significant Events are events that meet specific NRC criteria, including degradation of safety equipment, a reactor scram with complications, an unexpected response to a transient, or degradation of a fuel or pressure boundary Significant events are identified by NRC staff through detailed screening and evaluation of operating experience All WANO member nuclear plants must aspire to the following human performance objective; "The behaviors of all personnel result in safe and reliable station operation Behaviors that contribute to excellence in human performance are reinforced to continuously strive for event-free station operations" [2] V ERROR PREVENTION TECHNIQUES & BARRIERS In order to understand which error prevention techniques are most applicable, one must first understand what enablers can contribute to errors The criteria contained within this performance objective are assessed during peer reviews and its effectiveness reported There are two Nuclear Plant Event (NPE) definitions associated with human performance 12 main error enablers were identified and focused on as shown in Table I [7] - NPE08, “Human error which degraded nuclear safety related systems” - NPE09, “Human error which could have degraded nuclear safety related systems” Table I The Error Enablers If you look at the timeframe of when human performance error prevention was introduced and concentrate on the years 1992 to 2006, it is interesting to see the reduction in events at U.S nuclear plants This is shown in Fig [6] Time Pressure Distractions/Interruptions Fatigue/High workload Inexperience/Lack of knowledge Complacency Poor communication Stress Lack of assertiveness Resource planning Lack of Teamwork Lack of awareness Norms Plant trip risk procedures were assessed and each error enabler considered for the current task Suitable barriers were then applied and reviewed in action Barriers There are many barriers to prevent things from going wrong, they can be Organisational, Procedural and Physical The most important aspect is all barriers set by management are reinforced at every opportunity It would be their expectation for HUMAN PERFORMANCE IN THE NUCLEAR INDUSTRY staff to adhere to procedural usage, encouraged to have a questioning attitude and to stop when they are unsure Organisational defined The critical tasks and each step identified The work instructions and procedures are verified and common understanding checked This barrier use may be mandatory depending on the task The organisational barriers are the ones embedded within the company’s systems This makes it less likely that a plant modification occurs without drawing changes being in place coupled with operational and maintenance procedures There are many interconnected systems that will not allow the next step to take place until it is satisfied all the key elements to a successful outcome are met This cascades down to the competency levels of the person writing the work order instruction Using prior knowledge, operational or maintenance can be utilised at this point It demonstrates we are prepared to learn from past experience and use it effectively Prior knowledge can be in database format or personal experience Whatever method is used, it should capture previous incidents and near misses Stop, Think, Act, Review (S.T.A.R.) or Take / minutes to assess the work area are part of the self checking barrier This can be formalised by filling in a check sheet to demonstrate its use Confirmed communications is essential use at this point, to ensure the correct plant item is worked on The organisational barriers can contain latent errors These are hidden deficiencies in the process or values that provoke an error or cause the defense to break down The organisation also influences the culture at its locations through the reinforcement of its standards and expectations People are encouraged to work in a blame free culture but not to the extent where they are unaccountable for their actions One of the main organisational barriers which sets the benchmark for all expectations is training Shortfalls in training or a lack of training reduces the effectiveness of the understanding of what is required It is evident the individual plays a major part in effectively utilising the barriers If they have not taken personal ownership of the process and endeavor to use it, there is scope for errors occurring When people work around these barriers there is scope for error Work – The barriers used at this point can contain mandatory actions, depending on the work instruction Mandatory actions typically occur during the verification practices such as Peer Checking, Independent Verification or Concurrent / Simultaneous Verification Confirmed communications is also crucial during the work to exchange the right information at the right time Place keeping is another specific barrier employed during critical tasks to ensure the correct action is made at the right step Task Observations are carried when work is taking place This is an opportunity to carry out a formal or informal review of the complete scope of works It is a business improvement tool, used to capture the safety culture Procedural There are many procedural barriers in common use across industry They hold the individual responsible for their use The following typical work task and barriers used will highlight possible areas for concern A work task can be broken down into areas; Pre-work, Work and Post work Pre-work – The barrier used at this point is the Pre-Job Briefing Pre-work discussions are carried out when there is potential to impact on safety Everyone associated with the work is involved The roles and responsibilities are STEVEN M KONCZ surrounding the task A formal study of the work process also checks the standards & expectations are being met behaviour of individuals It is this behaviour which drives them to implement the error prevention tools or choose not to utilise them Self ownership of the processes and methodologies employed to prevent error are essential Observing these behaviors can take place at the point of work or checked remotely through documented evidence of the barrier being used Post work – This is an area where a PostJob Review takes place to determine if there are any areas for improvement or worthy of note for the next time Using this barrier enhances the operating / maintenance experience data gathering and can lead to further training, where appropriate It is also a documented opportunity to facilitate continuous improvement processes If we look at the point of work risk assessment Take 2, which encourages the person to take two minutes and review the potentials for error, the documented evidence can take the form of a tick sheet This barrier is open to any one of the error precursors stopping it from taking place, such as time pressure, complacency or high workload If no one double checks it took place, it could lead to an event Adding in an error prevention tool such as Independent Verification (I.V.), would make this process more robust It would only lengthen the risk assessment time slightly and possibly take three minutes with independent verification taking place or Take and I.V Although this could depend upon the working party numbers, it could be planned into the work pack This is an example of behaviour being observed and an additional barrier put in place Physical Physical barriers are the ones which prevent entry to areas that require specific access permissions The permit for work system is the procedural aspect that controls this type of barrier Boundary enclosures and containment buildings fall into this category also All of the barriers discussed were utilised in specific ways in the British Energy, Human Performance Awareness Workshops Similar barriers are used in WANO member nuclear power facilities, they are shown in Table II [7] Table II Error Prevention Tools Pre-Job Briefing Use of Operating Experience Procedural Use and Adherence Self checking (S.T.A.R.) Questioning Attitude (Stop When Unsure) Peer-Checking Independent Verification Clear Communication Techniques Post-Job Brief Task Observation Since people choose their behaviour at any given time, it is perhaps worth using the questioning attitude barrier but applying it to oneself prior to engagement with the task A prompt to make the person think how their behaviour will affect the task A very simple example is will I rush this job if I start it 30 minutes from meal time or end of shift? If a behaviour check is covered before a critical task, it may lead to the understanding that they could be distracted due to a personal issue playing on their mind Carrying out a formal self behaviour check is another way to enhance the error prevention process VI ERROR PREVENTION THE NEXT STEP It is well recognised that human performance error prevention hinges on the In the age of personal data devices and WiFi interconnectivity, there is now scope for HUMAN PERFORMANCE IN THE NUCLEAR INDUSTRY central databases with operating experience and error prevention tools to be available at the point of work, hazardous areas obviously excluded To avoid complacency with the known error prevention tools in use, revisiting all methodologies used and looking for ways to improve are advised Reviewing when things go right as well as wrong should also be trended to capture good practices for replication VII CONCLUSION Management commitment to focus on human performance, in particular error prevention and effective incident barriers, were the catalysts to improvements in this area Through external peer reviews and benchmarking current best practices, the UK nuclear industry took a collaborative approach to bring their power stations up to the expected standards They continue to maintain those standards and strive to exceed expectations ACKNOWLEGEMENTS The author would like to thank Ms Liesa Platten, of Synergy, Perth and Mr Joe Wade Human Performance Engineering Pty Ltd, Mandurah for independent verification of the readability of this document REFERENCES [1] Institute of Nuclear Power Operations, Excellence in Human Performance, INPO, Atlanta, 1997 [2] World Association of Nuclear Operators, Principles for Excellence in Human Performance, WANO-GL 2002-02, 2002 [3] The ARUP Journal 1/2006 Table 2, pg 15, 2006 [4] International Atomic Energy Agency, Managing Human Performance to Improve Nuclear Facility Operation, No NG-T-2.7, pg 1, 2013 [5] Practical Thinking, Edward de Bono, pg 11- 17, 1971 [6] Nuclear Regulatory Committee (NRC) Information Digest, 2006 [7] British Energy Group PLC, Human Performance Awareness Workshops, 2003 There are select businesses which invest directly in their staff by focusing on their innate human ability to make mistakes and how to take steps to prevent them from occurring Within a rational, unified, goal-seeking organisation, business improvement must have an understanding of human performance It is this understanding that can lead to improved business operations Trending of human performance errors should form part of the key performance indicators (KPI’s) This data can be derived from a robust route cause analysis process, which is performed by suitable qualified experienced persons Refreshing and repackaging the use of the error prevention tools, is essential for the success of the process and also facilitates continuous improvement Readdressing how the barriers are used in particular situations can contribute to the As Low As Reasonably Practicable (ALARP), process A formal behaviour self check, will make people think of additional barriers to use dependent upon how they may feel on the day Only they truly know what is going on in their own mind Nuclear Science and Technology, Vol.5, No (2015), pp.07-14 Codes for NPP severe accident simulation: development, validation and applications ARKADIY E KISELEV Nuclear Safety Institute of the Russian Academy of Sciences, B Tulskaya 52, 115191 Moscow, Russia E-mail: ksv@ibrae.ac.ru (Received 05 October 2015, accepted 25 October 2015) Abstract: The software tools that describe various safety aspects of NPP with VVER reactor have been developed at the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN) Functionally, the codes can be divided into two groups: the calculation codes that describe separate elements of NPP equipment and/or a group of processes and integrated software systems that allow solving the tasks of the NPP safety assessment in coupled formulation In particular, IBRAE RAN in cooperation with the nuclear industry organizations has developed the integrated software package SOCRAT designed to analyze the behavior of NPP with VVER at various stages of beyond-designbasis accidents, including the stages of reactor core degradation and long-term melt retention in a core catcher The general information about development, validation and applications of SOCRAT code is presented and discussed in the paper Keywords: nuclear power plant, safety, calculation codes, severe accident A long-term IBRAE RAN experience in developing software has allowed formulating the methodological approach that includes the following basic directions: - Working out the models based on the equations of mathematical physics and modern knowledge of processes and the phenomena that occur at various operating modes of reactor installations; - Based on these models, development and validation of the deterministic computer codes for nuclear power plants safety assessment; - Calculation-based and theoretical works to support the experimental programs; - Application of the developed software complexes for safety analysis of nuclear power plants Using the physical approaches to develop models allows considerable improving of the process and phenomenon modelling quality and reducing uncertainty of calculation results The software efficiency is tested through validation against experimental data While doing this, the assessment of the existing knowledge base on physical processes and phenomena is being conducted that allows formulating the tasks for experimental studies more accurately Participation of IBRAE RAS in the integral experiments is of a special significance since it allows verifying the new physical and numerical models in self-consistent way While developing models and codes for severe accident analysis, the basic uncertainties of the used physical models have been revealed; estimations of applicability of the existing codes to the safety analysis of NPP with various reactors have been made Most of IBRAE RAS codes are developed within the frameworks of joint projects with the Russian and international nuclear stakeholders ©2015 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute CODES FOR NPP SEVERE ACCIDENT SIMULATION… Table I Codes developed at IBRAE RIAN in cooperation with Russian and foreign institutions Emergency stage Basic IBRAE RAN codes Partners Early stage of reactor core degradation SVECHA, QUENCH, MFPR NRC/IPSN/EC FZK/RIAR Late stage of reactor core degradation CONV2D&3D LOHEY OECD/RRC KI/ IPSN Interaction of melt with concrete and catcher RASPLAV/SPREAD SPbAEP, NITI, AEP CONT REA/AEP, NRC/DOE Containment mechanics Thus, in the late 1990s, the works on development of the Russian code for safety analysis of new designs of NPP with VVER in conditions of severe accidents were started upon the initiative of JSC "SPbAEP" in cooperation of expert teams from IBRAE RAN, Russian Federal Nuclear Center “All-Russian Research Institute of Experimental Physics” (FSUE RFNC VNIIEF) and National Research Center «Kurchatov Institute» (NRC KI) Later, this code, which received the name SOCRAT, started to be applied also for safety assessment of the VVER projects operated or constructed in Russia In 2010, the basic version of the code SOCRAT was certified by the Russian regulator (Rostehnadzor) Since 2011, the work has been conducted on developing and validation of the advanced version of the code that allows assessments of the radiological consequences of severe accidents The quality of the models and validation allow considering the SOCRAT as a best-estimate code Integration of numerous physical models into one code provides end-to-end modelling of all essential stages of severe accidents and obtaining of the entire picture of the accident evolution from a moment of its occurrence (initiating event) up to release of radioactive fission products out of the NPP containment into the environment Thermohydraulic models of the integrated code SOCRAT describe the behavior of the two-phase coolant with non-condensable gases in the core, primary and secondary circuits of a reactor installation at all stages of severe accident including stage of total core uncover and stage of in-vessel melt retention They include the various modes of coolant flow, interphase interactions, various modes of heat exchange with walls of hydraulic channels, friction at channel walls, presence of noncondensable gases, coolant ejection under containment Also, the models of the SOCRAT code allow describing the operation of pumps, valves, hydraulic reservoirs and other elements of reactor installation equipment The set of the basic elements used to model the input deck of the primary and secondary circuits, allows describing the tracing of any hydraulic loops with the accuracy that is sufficient for modern calculations of severe accidents Thermohydraulic processes in a system of communicating containment rooms in are modelled self-consistently using the integrated in SOCRAT containment codes KUPOL-M and ANGAR, representing the certified codes with lumped parameters Physical mutually-consistent models describing the processes of fuel cladding ARKADIY E KISELEV oxidation by steam, thermomechanical behavior of fuel rods and absorbers, melting of reactor core and other in-vessel materials, melt relocation are used for numerical analysis of severe accidents at a stage of reactor core degradation While doing this, the real material composition of the reactor core is being taken into account The basic NPP objects that are modelled by the code SOCRAT in the advanced version are presented in Fig They are as follows: - Fuel; - Fuel assemblies; - Reactor core and in-vessel structures; - Reactor coolant system including safety systems; Code SOCRAT allows modelling the processes of melt interaction with water at a stage of melt retention in the lower plenum, formation and distribution of a corium liquid phase, stratification of metal and oxide components, reactor pressure vessel degradation and melt release into containment - Steam generator and main steam line; - Containment Fig Phenomena modeled in SOCRAT CODES FOR NPP SEVERE ACCIDENT SIMULATION… Fig Main processes simulated in the layer of homogeneous melt The basic physical models of the integrated code SOCRAT at in-vessel stage of accident are presented below: The experimental programs that were used for the code validation are as follows: CORA, QUENCH (Germany), PHEBUS (France), RASPLAV, MASCA (Russia OECD), ISTC/PARAMETER, ERCOSAMSAMARA (joint Rosatom-Euroatom project), LOFT, PBF, international standard problem ICSP MASLWR, international benchmark BSAF (analysis of the accident at the Fukushima Daiichi NPP) The advanced version of the integrated code SOCRAT allows carrying out calculations of parameters required for assessing the radiological consequences of severe beyond design-basis accidents at NPP with VVER reactors and, in addition to the basic version, describes in details the following processes: - Buid-up of radioactive fission products (FP) in fuel and their release into the fuel rod’s gas gap; Fig shows the calculated and measured temperatures of the surface of the fuel assembly simulator in the PARAMETER/SF1 experiment The PARAMETER program investigates phenomena associated with reflooding of a degrading VVER like core under postulated severe accident conditions, in an early phase when the geometry is still mainly intact The figure confirms that the SOCRAT code correctly reflects the dynamics of the fuel assembly temperature behavior at all stages of the experiment (heating up, oxidation, and overheated core re-flood) under conditions of the presence of chemical power sources and convection and radiation heat exchange This results from a sufficiently large set of models of SOCRAT code and their validation in a wide range of initial data - Transport and sedimentation of radioactive fission products in various physical and chemical forms in the reactor primary circuit and in the containment; - Release of radioactive fission products into environment Permanent validation of the SOCRAT code as well as of its physical models is one of the most important stages of the development and application Models and algorithms of the SOCRAT code have passed all-round assessment against large data set, received in separate effect tests and integral experiments performed in Russia and abroad 10 STUDY ON TREATMENT OF RADIOACTIVE LIQUID WASTE FROM URANIUM ORE PROCESSING … Table IV Compositions of liquid waste after Preliminary precipitate and deeply treatment by nano magnetite material Number Elements Concentration Unit Analysis method Fe 0.542 mg/l ICP-MS Al 0.66 mg/l ICP-MS Cr 0.001 mg/l ICP-MS Mn 0.009 mg/l ICP-MS Mg 0.23 mg/l ICP-MS Zn 0.025 mg/l ICP-MS As 0.006 mg/l ICP-MS U 0.014 mg/l ICP-MS Total activity α 0.097 Bq/l MPC-2000 measure total activity α,β 10 Total activity β 0.985 Bq/l MPC-2000 measure total activity α,β Comments: The data from Table IV revealed that liquid waste from uranium ore processing after preliminary precipitate and deeply treatment by nano ferromagnetic material was reached discharge environment standard of QCVN 40:2011 The research team examined the adsorptive capacity of different materials and tested them with liquid waste from uranium ore processing wich was preliminary precipitated at pH The results are given in the following table: C Comparing the applicability of different materials for the liquid waste treatment of uranium ore processing Table IV Adsorption capacity of differently materials Number Materials Type Particular size (nm) Specific surface area (m2/g) Adsorption capacity (mg/g) Fe3O4 Vietnam 80 - 100 50 - 70 53.5 NiFe2O4 Vietnam 70 - 90 60 - 80 58.5 Fe3O4 Slovakia 20 - 30 100 - 110 82.2 NiFe2O4 Slovakia 15 - 20 110 - 120 86.5 Table V Compare the results of treatment of categories materials Number Materials Type Weight materials (g) The total activity of α (Bq/l) The total activity of β (Bq/l) Fe3O4 Vietnam 0.097 0.985 NiFe2O4 Vietnam 0.089 0.865 Fe3O4 Slovakia 0.5 0.092 0.95 NiFe2O4 Slovakia 0.5 0.085 0.876 44 VUONG HUU ANH et al Comment: From the data shows a comparison between two kinds of Vietnam and two types of Czechs that these of Czechs have higher usability But the main solution is the cost of this kind almost double the price of Vietnamese types So, the material Vietnamese NiFe2O4 was selected for application to handle liquid waste for uranium ore processing from uranium solution and uranium ore processing uranium was assessed Propose process treatment liquid waste from uranium ore processing with the application of nano oxide ferromagnetic was submited From an examination of the affective parameters to the adsorption of nano oxide ferromagnetic for uranium solution and test for real liquid waste of uranium ore processing We have a proposed process using of nano oxide ferromagnetic to handle liquid waste from the of uranium ore processing D Proposed technology process for the application of nano oxide ferromagnetic materials to treat liquid radioactive waste from processing of uranium ores The applicability of nano oxide ferromagnetic material for adsorption uranium Liquid Wate pH = 2.2 NaOH Preliminary precipitation pH = 8.0 Waste sludge Divide solid – liquid and flush Nano oxide magnetite (NiFe2O4) Adsorption by nano oxide magnetite Waste sludge Divide solid – liquid Wastewater Liquid no satisfy standard Cement of sludge Analysis test Crating temporary storage Discharged Disposal near the surface Fig.4 Technology process for treatment liquid radioactive waste generated from uranium ores processing 45 STUDY ON TREATMENT OF RADIOACTIVE LIQUID WASTE FROM URANIUM ORE PROCESSING … CONCLUSIONS REFERENCES [1] Cao Hung Thai, “Introduction fuel cycle, management and disposal of radioactive waste”, Institute for Technology of Radioactive and Rare Elements, 2006 The effective parameters of adsorption processing using nano oxide ferromagnetic was tested with simulated uranium solution: temperature, stirring rate, stirring time, the pH value of initial uranium concentration [2] Nguyen Dinh Trieu, “Documentation of physics analytical methods and physical chemistry” [3] Than Van Lien, “Hydrometallurgical uranium”, Publisher of Vietnam National University, 2004 The radioactive elements adsorption capacity of nano oxide ferromagnetics were applied onreal liquid waste solution from uranium ore hydrometallurgical processing After treatment, treated liquid waste reached the discharge criteria to the environmental of QCVN40:2011 [4] National technical Regulation on industrial wastewater, Ministry of Natural Resources and Environment of the Socialist Republic of Vietnam, QCVN 40:2011 [5] Standard 4397:1987 RECOMMENDATION [6] James D Navratil, Archive of Oncology; 9(4):257-60 “Advances in treatment methods for uranium contaminated soil and water”, 2001 Due to the conditions and duration of the study are limited, the theme did not studied completely the adsorption of uranium from waste solution by nano oxide ferromagnetic [7] James D Navratil, Vol (5): 697-702, "Chemistry of Iron Ferrites and Their Application for Wastewater and Acid Mine Water Treatment”, 2010 The theme did not study elution processing regeneration nano oxide ferromagnetic after adsorption and have not survey the continuous adsorption column method [8] James D Navratil and Andrew C.Akin, “Mine water treatment using magnetite and iron ferrites”, 2009 Thus, in the next time, the theme need to research uranium adsorption with nano oxide ferromagnetic and desorption processing The adsorption of nano oxide ferromagnetic for the radioactive nuclides and other heavy metals were expanding studied of in liquid waste 46 Nuclear Science and Technology, Vol.5, No (2015), pp 47-58 Verification of TVS-2006 fuel rod design of VVER-AES2006 reactor under steady-state operating condition Using FRAPCON-3.5 code Dinh Van Chien Vietnam Atomic Energy Agency, 113 Tran Duy Hung, Cau Giay, Hanoi, Vietnam Email: dvchien@most.gov.vn (Received 10 August 2015, accepted 22 September 2015) Abstract: The purpose of this paper is to discuss the independent verification of TVS-2006 fuel rod design used in VVER-AES2006 reactor (Novovoronezh NPP-2 Power, Unit 1), based on the acceptance criteria and the reference data given in the Preliminary Safety Analysis Report of the State Research, Design, Construction and Survey Institute “Atomenergoproekt” (PSAR) and the operation of VVER-1000 reactor The calculations were performed using FRAPCON-3.5 code, including fuel temperature, cladding temperature, fission gas release, internal gas pressure, cladding stress and strain, fuel extension, fuel rod elongation, cladding creep rate, fuel swelling rate, cladding oxide thickness and hydrogen concentration The results are compared with the calculated data using START-3 code in PSAR and the acceptance criteria required by Russian nuclear regulatory body Despite some discrepancies, the results showed conformance with the calculated data given in the PSAR and meet the acceptance criteria Keywords: Nuclear fuel rod design, fuel behaviour, design verification, acceptance criteria I INTRODUCTION From the 80s of the 20 th century to the present, the fuel rod design has been continuously improved to optimize the fuel rod behaviour and meet the higher operating conditions of reactors, such as the high-power level (1000-1600 MWe), power uprate up to 110%, increased burn-up (60-70 GWd/tU) and extended fuel cycles (from 12 to 18 months) Thus, more realistic predictions of fuel performance is needed to allow operating Nuclear Power Plant effectively and safely, as well as improving operating margins and efficiency and higher flexibility in fuel management So, a reliable prediction of fuel rod behaviours is important for fuel rod design and safety evaluation in nuclear power reactors [1] While the fuel rod design is performed by the vendors using their own codes, the utilities and the safety authorities also need to perform independent design verification using licensing fuel rod codes such as FRAPCON-3.5, COPERNIC, TREQ, PAD codes FRAPCON-3.5 code [2, 3], one of fuel performance codes verified and licensed by United States Nuclear Regulatory Commission (US.NRC) to review fuel design of Light Water Reactor (LWR), is designed to perform the thermal-mechanical calculations of LWR fuel rod such as the temperature, pressure, and deformation as functions of time-dependent fuel rod power and coolant boundary conditions under steady-state condition, and to generate initial conditions for transient fuel rod analysis using the FRAPTRAN-1.5 code [4, 5] The FRAPCON- ©2015 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute VERIFICATION OF TVS-2006 FUEL ROD DESIGN OF VVER-AES2006 REACTOR… 3.5 code uses data of material properties documented in the updated version of the MATPRO material properties package for high burn-up conditions and advanced cladding alloy such as Zircaloy-2, Zircaloy-4, ZIRLOTM, M5, The main models of FRAPCON-3.5 code used in the calculations include the FRACAS-I thermal-mechanical model, Forsberg-Massih fission gas release model and Cladding oxidation and hydrogen content models II CALCULATION MODEL FOR TVS2006 FUEL ROD A Description of TVS-2006 fuel rod design The TVS-2006 fuel rod design of VVERAES006 reactor is developed on the basis of TVS-2 fuel rod design, which has been developed by EDB “Gidropress” FSUE for the commercial VVER-1000 reactor, using the design solutions, calculations as well as the experimental justification A TVS-2006 fuel rod comprises the following parts: Upper plug, cladding, lower plug, fuel pellets and a spring (Fig 1, Table I) [6] Until now many features of FRAPCON-3.5 code have been improved to be used in the independent review and safety analysis of fuel rod design, as well as in the operational and licensing supports by some authorities as US.NRC, AREVA NP, Inc (USA), IRSN (France), ALVEL, NRIR (Czech Republic), CRCD (Ukraine), KEPCO (Korea), NRA (Japan), Tractebel Engineering S.A (Belgium)…Although some calculations of fuel rod behaviour of VVER-440/VVER-1000 reactors have been performed by FRAPCON-3.5 code but not yet applied for fuel rod of VVER-1200 reactor (VVER-AES006) Therefore, the independent verification of fuel rod design of VVER- AES006 reactor (Novovoronezh NPP-2 Power Unit 1), TVS-2006 fuel rod [6], has been chosen as the main objective of this research The obtained results, using the FRAPCON-3.5 code, are also compared with the calculated data by START-3 code in PSAR and the acceptance criteria In which, the acceptance criteria using in evaluation are usually established by the fuel vendors based on experimental data and theoretical considerations, with adequate margins that are accepted by the Russian Safety Authority Fig Configuration of TVS-2006 fuel rod 48 DINH VAN CHIEN Table I Main parameters of TVS-2006 fuel rod Parameter Value Number of fuel rods in fuel assembly 312 Fuel rod lattice Evenly triangular Fuel rods pitch, mm 12.75 Fuel UO2 Fuel density, kg/m3 (10.4-10.7).103 Mass fraction of uranium isotopes mixture in fuel, % Cladding material ≥87.9 E110 (Zr-1%Nb) Total fuel rod length, mm 4033 Outer diameter of fuel cladding, mm 9.10 ± 0.04 Inner diameter of fuel cladding, mm 7.73 ± 0.06 Outer diameter of fuel pellet, mm 7.60 ± 0.03 Diameter of centreline hole in fuel pellet of a fuel rod, mm Grain size in fuel pellet, µm 1.2 ± 0.2 10-20 Fuel pellet height, mm 9.0-12.0 Fuel column height (cold state), mm 3730 Fuel mass in fuel rod, kg 1.712 Average linear power, W/cm 167.8 Peak linear power, W/cm 420 o Maximum cladding temperature, C 235 Enrichment U 355 (maximum value), % 4.95 ± 0.05 The cladding of TVS-2006 fuel rod is made of E110 alloy, however, properties of E110 alloy is not modelled in FRAPCON-3.5 code Therefore, M5™ alloy is selected instead of E110 because it has the similar chemical composition as E110 alloy (Zr-1%Nb) [7]; B Modelling method The TVS-2006 fuel rod has been modelled using the FRAPCON-3.5 code based on the design parameters, reference data in the operation of VVER-1000 reactor [3, 5, 7, 8, 9, 10, 11] and the given data of PSAR [6]: Calculations have been performed for fuel cycles, the length of each cycle is 343.2 Effective Full Power Days (EFPD), using the power history taken from given data of PSAR and reference data in the operation at VVER1000 reactor (Fig 2) [6, 7, 10, 11] Conservative axial power distribution in the hottest fuel rod was used for evaluating maximum temperature of fuel and cladding (Table III) [6] The dimensions for TVS-2006 fuel rod were taken from design data The fuel rod was divided into 50, 17 and 45 for number of equallength axial nodes, radial boundaries in the pellet and equal-volume radial rings, respectively; The initial fill pressure of fuel rod, coolant pressure, coolant inlet temperature, and mass flux of coolant were all taken from design data (Table II) [6] 49 VERIFICATION OF TVS-2006 FUEL ROD DESIGN OF VVER-AES2006 REACTOR… Fig The linear heat generation rate of the fuel rod during operation Table II Main parameters of the boundary conditions Parameter Value The rod initial fill pressure, MPa 2.1 Coolant system pressure, MPa 16.2 ± 0.3 Coolant inlet temperature, oC 298.2 ± Mass flux of coolant, kg/s.m2 3930 Table III Conservative axial power distribution (Kz) in the hottest fuel rod Parameter Core height, % Kz Value 15 25 35 45 50 55 65 75 85 95 0.5 0.83 1.07 1.25 1.35 1.357 1.35 1.25 1.07 0.83 0.5 design parameter with nominal input data and best estimate models are performed by FRAPCON-3.5 code Then, the assessment of design parameter sensitivities and quantification of design uncertainties via the Root Mean Square method are used to calculate the maximum values of the calculation results, to be compared with the acceptance criteria and the calculated data by START-3 code in PSAR [6, 12] C Design verification method The verification of fuel rod design is performed by the deterministic calculations using FRAPCON-3.5 code and also combines with uncertainty evaluation by the statistical method using the Root Mean Square The sources of uncertainties include operation parameters (power histories, flow rate, pressure…), manufacturing parameters (cladding thickness, pellet diameter, fuel density…) and key correlations (fuel thermal conductivity, densification, swelling, fission gas release, cladding creep, corrosion, hydrogen pickup) Best estimate calculations of the III CALCULATION RESULTS A Thermal-mechanical calculation results 50 DINH VAN CHIEN interaction, stress corrosion cracking… So, the excessive fission gas release can cause the rod pressure to rise beyond system pressure and lead to fuel damage Thus, rod pressure need to be limited by safety criteria and must be calculated for the design evaluation The results of thermal-mechanical calculations are given in Table and Figs 3-6 (nominal values), including: Fuel temperature, cladding temperature, fission gas release, and internal gas pressure The fuel rod temperature distribution depends on design parameters, materials properties and on many phenomena which develop during irradiation Many properties are exponentially dependent on temperature Maximum fission gas release of fuel rod (FGR) is 3.58% at the end of 4th cycle and close to calculated FGR by START-3 code (~3%) Maximum rod internal pressure is 5.69MPa during four cycles of operation with the safety margin K = 2.85 The calculation results of FGR and internal pressure show the guarantee of design in order to protect the fuel against cladding lift-off These results are lower than the limit values and show that they ensure to prevent the diametral gap between the fuel and the cladding from re-opening during steadystate operation, which causes ballooning and affect the coolant flow or the local overheating of the cladding The results of fuel temperature calculations show that the fuel temperature (Tf) reaches its maximum Tfmax = 1746.37K at the beginning of the first cycle of operation and is lower than the limit value [T] = Tmelt = 3113.14K with the safety margin K = 1.78 The maximum of average fuel temperature in four cycles is 1020.25K For cladding temperature, the maximum cladding outside temperature (Tc) is 625.19K at the beginning and at the end of the first cycle of the operation and does not exceed limit value of 628.15K These values are close to calculated data by START-3 code in PSAR (maximum fuel and cladding outside temperature are 1860.15K and 627.25K, respectively), also meet acceptance criteria and protect the fuel against excessive degradation of cladding mechanical properties related to hydrogen pickup or accelerated oxidation (high cladding surface temperatures) The calculation results of fuel rod temperature show the guarantee of design in order to protect the fuel against any types of failures resulting from fuel melting or overheating Therefore, accurate temperature estimates are important for many safety design criteria As above analyses, the thermal-mechanical calculations have demonstrated that the results are clove to calculated data by START-3 code in PSAR, satisfy acceptance criteria and also show adequate thermal-mechanical reliability of TVS2006 fuel rod in operation However, the rod internal pressure is quite low, which the reason may be due to the insufficient information of the design power histories and the modelling method by START-3 code It was found that the average variance in rod internal pressure value for a biased similar PWR fuel rod by adjusting the steady-state power by ±10% is 20% Also, the average value of rod internal pressure varied by approximately 32% when the fuel thermal conductivity model is biased by ±0.5W/m-K [12] Fission gas release (FGR) and rod internal pressure (Pi) have a major impact on mechanical properties of fuel rod Fission gas release can cause fuel swelling, pressure buildup (xenon, krypton), pellet-cladding mechanical 51 VERIFICATION OF TVS-2006 FUEL ROD DESIGN OF VVER-AES2006 REACTOR… Besides, as shown in Table 4, the internal pressure presented in the PSAR seems to be too high with regard to experience feedback and with longer fuel length in the design, it must be around 10 to 12 MPa after four cycles of irradiation instead of 15.2 MPa (from operational feedback measurements and calculations on irradiated fuel rods) [12,13] Thus, it should be verified with additional calculations with appropriate power history Fig Fuel centreline temperature Fig Cladding outside temperature Fig Fuel centreline temperature Fig Fission gas release Fig Rod internal pressure Table IV Results of thermal-mechanical calculations Parameter Nominal Results Uncertainty Maximum Results PSAR data Deviation, % Limit value Safety margin, K* Standard safety margin, [K] 1551.7 194.67 1746.37 1860.15 -6.12 3113.14 1.78 1.1 623.38 1.81 625.19 627.25 -0.33 628.15 1.005 - FGR, % 2.44 1.14 3.58 ~3 - - - - Pi, MPa 4.97 0.72 5.69 15.2 -167.14 16.2 2.85 1.1 Tf, K Tc , K *Safety margin K = Limit value/Maximum value 52 DINH VAN CHIEN Although the result meets design criteria but it is quite low than calculated data in PSAR (190.4 MPa) because E110 alloy is modelled exactly by FRAPCON-3.5 code and for the calculations, M5™ alloy was assumed since it has the same chemical composition as E110 alloy (Zr-1%Nb) [7] B Strength calculation results The strength calculation results of cladding are given in Table and Figs 7-10 (nominal value) The operating experience of fuel rods as well as calculations and experiments show that hoop stress and strain determine cladding strength in steady-state conditions and during transients, that is why they will be in the focus of further strength analysis For the cladding hoop strain (εh), the maximum value is 0.20% with the safety margin K = 2.5 For the cladding elastic strain, the maximum values are 0.06%, 0.09% and 0.0629% for hoop strain (εeh), axial strain (εea) and radial strain (εer), respectively These results of strain show ability to protect the fuel against pellet-cladding interaction (PCI) failure The intent of these analyses is to ensure integrity of cladding due to slow rate strain accumulation at which the stress does not reach the stress limit (yield stress) The calculation results of stress and strain show that they satisfy acceptance criteria and steady-state operating conditions However, it is noticed that the stress is not an adequate criterion for fuel failure since it cannot be measured during irradiation and these calculations should be considered in the Condition II transients of ramp power In the beginning of the cycle, hoop stress on the internal cladding surface are mainly determined by thermal gradient and external differential pressure After closure of the radial gap between fuel and cladding, the fuel first comes into “soft” contact with cladding, and the contact becomes “hard” after crack healing in fuel As a result, the hoop stress of the internal cladding surface increase first in the central and then in the side cross-sections of fuels rods During four cycles of operation, the stress reaches a steady level of about 70-80 MPa and the maximum effective cladding stress (σeff) is 103.28 MPa This value is lower than the yield stress of cladding material in irradiation conditions (340-350 MPa) The maximum cladding hoop stress (σh) is 95.49 MPa Fig Cladding hoop stress Fig Effective cladding stress 53 VERIFICATION OF TVS-2006 FUEL ROD DESIGN OF VVER-AES2006 REACTOR… Fig Cladding hoop strain Fig 10 Cladding elastic hoop strain Table V Results of strength calculations Parameter Nominal Results Uncertainty Maximum Results PSAR data Deviation, % Limit value Safety margin, K* Standard safety margin, [K] σeff, MPa 83.32 19.96 103.28 - - 350 3.39 - σh, MPa 76.99 18.50 95.49 190.4 -49.8 230 2.41 1.2 εh , % 0.14 0.06 0.20 - - 0.5 2.5 - εeh, % 0.05 0.01 0.06 - - - - - εea, % 0.07 0.02 0.09 - - - - - εer, % 0.06 0.0029 0.0629 - - - - - Table VI Results of deformation calculations Parameter Nominal Results Uncertainty Maximum Results PSAR data Deviation, % Limit value Safety margin, K* Standard safety margin, [K] ∆H, mm 3.18 1.84 5.02 - - - - - ∆L, mm 18.31 11.92 30.23 47.7 -36.62 61.6 2.04 - vcreep, 10-11 m/m/s 5.78 0.70 6.48 - - - - - vswell, 10-11 m/m/s 7.63 1.01 8.64 - - - - - 54 DINH VAN CHIEN C Deformation calculation results The results of deformation of fuel rod are given in Table and Figs 11-13 (nominal value), including: Fuel stack axial extension, fuel swelling rate, fuel rod elongation, and cladding creep rate Therefore, fuel cladding has to be ensured against cladding creep collapse (axial slip failures) The results show that the maximum cladding creep rate (vcreep) is 6.48 (10-11 m/m/s) and maximum fuel swelling rate (vswell) is 8.64 (10-11 m/m/s), this ensures no failure of cladding Fuel rod length changes due to irradiation effects and differential thermal expansion shall cause interference with the fuel assembly structure This evaluation is a critical design input because it determines the assembly length (fuel assembly mechanical design) The analyses have to show ability to prevent cladding from axial buckling or overstressing of the thimble tubes and/or thimble-to-nozzle connections Fig 11 Cladding axial extension The results show that during four cycles of operation, maximum fuel stack axial extension (∆H) is 5.02 mm, and maximum fuel rod elongation (∆L) in operating conditions is 30.23 mm The result meet design criterion but it is lower than calculated data in PSAR (47.7 mm) because in the calculations, M5™ alloy was assumed since it has the same chemical composition as E110 alloy (Zr-1%Nb) [7] Taking into consideration the thermal elongation and irradiation growth (about 0.15 %) of fuel assembly skeleton, the clearance between the upper fuel rod plugs and the fuel assembly head in a hot state is about 61.6 mm without fuel rod elongation This value is considered the limit value Thus, the calculated safety margin for fuel rod elongation is K = 2.04 Fig 12 Fuel swelling rate The cladding can be free standing at beginning of life (before densification) and no long-term buckling However, fuel cladding is prone to instant collapsing when reaching critical pressure for this cladding state as well as to long-term accumulation of creep deformations and fuel swelling Fig 13 Cladding creep rate 55 VERIFICATION OF TVS-2006 FUEL ROD DESIGN OF VVER-AES2006 REACTOR… D Cladding oxidation calculation results and The results of surface corrosion and cladding hydration calculation show that maximum oxide thickness is 20.21μm and maximum hydrogen concentration is 73.42 ppm with the calculated safety margins are 2.97 and 5.45 for surface corrosion and cladding hydration, respectively The results meet design criteria but it is lower than calculated data in PSAR (oxide thickness 30 μm) This deviation can be due to assuming M5™ alloy instead of E110 alloy [7] The calculation results have showed that cladding of TVS-2006 fuel rod can meet operating ability in normal condition of reactor for the cladding oxidation and hydration hydration The results of oxide thickness and hydrogen concentration of cladding are given in Table and Figs 14-15 (nominal value) Oxidation and hydriding under normal operating conditions of reactor directly impact fuel performance, not only during normal operation, but during transients and accidents as well Cladding corrosion reduces the effective thickness of the cladding, decreases the effective thermal conductivity of the cladding and thus increases the cladding and fuel temperatures and also reduces effective cladding-to-coolant heat transfer Hydrogen absorption by the cladding and subsequent formation of hydrides may lead to cladding embrittlement These phenomena are increasingly important at higher exposures So, the analyses have to show ability to protect the fuel against any type of cladding corrosion induced failure Fig 14 Cladding oxide thickness Fig 15 Claddinghydrogen concentration Table VII Results of oxide thickness and hydrogen concentration calculations Parameter Nominal Results Uncertainty Maximum Results PSAR data Deviation, % Limit value Safety margin, K* Standard safety margin, [K] Oxide thickness, μm 15.11 5.10 20.21 30 -32.6 60 2.97 1.5 Hydrogen concentration, ppm 68.05 5.37 73.42 60-80 - 400 5.45 - 56 DINH VAN CHIEN Additional, when the calculations have been performed on two computing systems of the Vietnam Atomic Energy Agency (VAEA) and Tractebel Engineering (TE, GDF Suez, Belgium) with the same input, it has been indicated that the calculation results are similar This shows the reliability and the realistic meaning of the used tools at VAEA such as computer server and FRAPCON-3.5 code version Also, this is the first calculation result using FRAPCON-3.5 code in order to verify the TVS-2006 fuel rod design of VVER-AES2006 reactor IV CONCLUSIONS The independent verification of the TVS2006 fuel rod design of VVER-AES2006 reactor under steady-state operating condition using FRAPCON-3.5 code was performed based on the design parameters and the reference data from the operation of the VVER1000 reactor The calculation results are compared with acceptance criteria and the obtained data by START-3 code in the PSAR It has been indicated that the calculation results show conformable tendency of the operating behaviors, as well as satisfying the operational ability in normal condition of reactor and are also close to obtained data by START-3 code in the PSAR ACKNOWLEDGEMENT This work is performed under the research framework of the national project hosted by the VAEA The author would like to thank the VAEA and the National Foundation for Science and Technology Development (NAFOSTED) for supporting administrative procedure and finance Also thanks to the TE, GDF Suez for cooperation in the training program in Brussels, Belgium, during AugustOctober 2014 Specially, the author would like to express my gratitude to Dr Hoang Anh Tuan, Dr Tran Dai Phuc at VAEA and Dr Jinzhao Zhang at TE, GDF Suez for their guidance and comments, as well as the colleagues for many discussions The calculation values by FRAPCON-3.5 code are lower than the limit values of acceptance criteria and safety margins are greater than standard safety margins.The deviations between calculation results of FRAPCON-3.5 code and START-3 code may be due to the insufficient information about the design power histories and the modelling method using START-3 code in the PSAR, as well as assuming M5™ alloy (similar Zr-1%Nb alloy) instead of E110 alloy due to FRAPCON3.5 code does not model exactly characteristics of E110 alloy However, it was found that the average variance in the calculation parameters for a biased similar PWR fuel rod by adjusting the steady-state power by ±10% is 20% 57 VERIFICATION OF TVS-2006 FUEL ROD DESIGN OF VVER-AES2006 REACTOR… [8] L Yegorova, “Data Base on the Behavior of High Burn-up Fuel Rods with Zr-1%Nb Cladding and UO2 Fuel (VVER Type) under Reactivity Accident Conditions”, NUREG/IA0156, Vol.1 Review of Research program and Analysis of Results, US NRC, 1999 REFERENCES [1] Jinzhao Zhang, Simulation of fuel behaviors under LOCA and RIA using FRAPTRAN code and uncertainty analysis with DAKOTA, IAEA Technical Meeting on Modeling of WaterCooled Fuel Including Design Basis and Severe Accidents, China, November, 2013 [9] L Yegorova, V asmolov et al., “Data Base on the Behavior of High Burn-up Fuel Rods with Zr-1%Nb Cladding and UO2 Fuel (VVER Type) under Reactivity Accident Conditions”, NUREG/IA-0156, Vol.2 Description of Test procedures and analytical Methods, US NRC, 1999 [2] K.J Geelhood, W.G Luscher and C.E Beyer, “FRAPCON-3.5: A Computer Code for the Calculation of Steady-State, ThermalMechanical Behaviour of Oxide Fuel Rods for High Burn-up”, NUREG/CR-7022, Vol.1, US NRC, 2014 [10] L Yegorova, G.Abyshov et al., “Data Base on the Behavior of High Burn-up Fuel Rods with Zr-1%Nb Cladding and UO2 Fuel (VVER Type) under Reactivity Accident Conditions”, NUREG/IA-0156, Vol.3 Test and calculation results, US NRC, 1999 [3] K.J Geelhood, W.G Luscher and C.E Beyer, “FRAPCON-3.5: Integral Assessment”, NUREG/CR-7022, Vol.2, US NRC, 2014 [4] K.J Geelhood, W.G Luscher and J.M Cuta, “FRAPTRAN-1.5: A Computer Code for the Transient Analysis of Oxide Fuel Rods”, NUREG/CR-7023, Vol.1, US NRC, 2014 [5] [11] Taylor S Blyth, “Fuel performance code benchmark for uncertainty analysis in light water reactor modeling”, USA, 2012 K.J Geelhood and W.G Luscher, “FRAPTRAN-1.5: Integral Assessment”, NUREG/CR-7023, Vol.2, US NRC, 2014 [12] K.J Geelhood, W.G Luscher and C.E Beyer, “Predictive Bias and Sensitivity in NRC Fuel Performance Codes”, NUREG/CR-7001, PNNL-17644, US NRC, 2009 [6] I.I Kopytov, S.B.Ryzhov, Yu.M Semchenkov et al., “Prelimary safety analysis report Novovoronezh NPP-2 Power Unit 1”, Rusia, 2009 [7] [13] FRAMATOME Evaluation du comportement thermomecanique du crayon combustible UO2, FF.DC-0012, 2003 A Shestopaalov, K Lioutov, L Yegorova, “Adaption of USNRC’s FRAPTRAN and IRSN’s SCANAIR Transient codes and Updating of MATPRO Package for Modeling of LOCA and RIA Validation Cases with Zr1%Nb (VVER type) Cladding”, NUREG/IA0209, US NRC, 2003 58 [...]... weighting functions for the dynamic rod worth simulation The reason for the evaluation of the AWFs and SAFs instead of the threedimensional weighting functions generated using Eq (3) is explained as follows The AWFs for the excore detector at RCP were illustrated in Fig 2 The relative differences of AWFs at RCP and at G 1IN, G 2IN, or G 3IN were provided in Figs 3-5 The SAFs at RCP and G 1IN, G 2IN, or G 3IN. .. gelatin increases the current efficiency and decreases the energy consumption Gelatin when present in the solution polarizes the cathode causing the electroreduction of cadmium at more negative potentials The presence of gelatin affects the degree of crystallinity of the electrodeposits indicating that the deposits are also more ductile Scanning electron micrographs of cadmium deposits obtained in the. .. uranium loading of the two resins tends to increase with increasing pH of solution Uranium loading of GS300 resin increases from 28.5 to 50.2 gU/l of wet resin when pH of solution increases from 1.2 to 1.8 Similarly, uranium loading of A400 resin increases from 27.1 to 47.7 gU/l However, in the pH range from 1.6 to 1.8 uranium loading of the two resins does not increase significantly In the same experimental... strong base anion resin Amberlite IRA-400 and IRA-420 These resins have high uranium capacity with good selectivity However, these resins are expensive Therefore, the trial of other ion exchange resins to avoid the dependence on one resin in the processing of solutions containing uranium is needed In this work, the authors will present the test results on two commercial resins Indion GS300 and Purolite... was fully inserted into the core while other control rods at RCP were determined and inter-compared instead of extremely large numbers of the calculated three-dimensional weighting functions The results indicate that the weighting functions were relatively insensitive to the control rods position during the rod drop experiments and consequently those weighting values at RCP can be applied in the dynamic... as at RCP and G 1IN Instead, the AWFs and SAFs at RCP and at G 1IN, G 2IN, or G 3IN, were determined and intercompared ∫ ∑ ∑ The AWF for the FA (i,j) which represents the detector response contributions from individual FAs is calculated by Eq (4) (3) where is the fission spectrum at energy group g and ∑ the adjoint flux at the FA (4) layer (i,j,k) at energy group g Thereafter, these weighting functions were... results, for the similar solution uranium loading of IRA-420 resin reached 55 gU/l at pH of 1.6 Thus, the uranium loading of A400 resin is only equal to 80% of IRA-420 resin, while uranium loading of GS300 resin was 85% 0.6 0.5 0.4 8 min 4 min 0.3 2 min 0.2 0.1 0 0 Because total capacity of IRA-420, GS300 and A400 resin is the same, so the difference in uranium loading of GS300 and A400 resin comparing to... concentration in effluent solution equals to the concentration of feeding solution After the soption phase, the resin layer is washed with a solution of sulfuric acid 1/1000 (volume) for separating off the feeding solution in the column Then uranium desorption was conducted by using a mixture of 1M of NaCl and 0.05M of H2SO4 [1, 2, 3] Uranium, iron and other impurities containing in the eluate solution... advantage of the Monte Carlo method is the capability of modeling reactor configurations with arbitrary geometrical complexity With the Monte Carlo method, one can also choose either the forward method or the adjoint method The Monte Carlo forward method allows the calculation of the weighting function value of a given point in the reactor and therefore gives more detailed results than the adjoint method... of the following operations: oxidation of cadmium; leaching; cleaning of the solution and precipitation of the cadmium sponge; oxidation of the sponge, its repeated dissolution and cleaning of the solution; electrowinning; smelting of cathodic cadmium [3, 4] Recovery of cadmium by hydrometallurgy combined with electrolysis is commonly used for the recovery of cadmium in the process of purifying zinc