Assessment of thermal aging embrittlement in a cast stainless steel valve and its effect on the structural integrity

This paper analyzes the thermal aging embrittlement occurred in a cast stainless steel valve, which is part of the reactor water clean-up (RWCU) system of a Spanish boiling water reactor (BWR) nuclear power plant. The aim is to estimate the current and future state of the material and the corresponding structural integrity of the valve. Given that there is no data available for the experimental characterization of the material, the evolution of the mechanical properties (fracture toughness, yield stress, flow stress and Ramberg-Osgood parameters) has been estimated using the ANL procedure.With the obtained estimations, the critical crack size has been calculated using the European procedure FITNET FFS and the ASME Code.This analysis considers not only the evolution of the mechanical properties up to now but also its future evolution in case there is a life extension of the plant until year 2029.     

Gas-cooled reactor thermal hydraulic analyses with MELCOR

Pursuant to the Energy Policy Act of 2005, the High Temperature Gas-Cooled Reactor (HTGR) has been selected as the reference design for the Next Generation Nuclear Plant (NGNP). Stemming from a U.S. Nuclear Regulatory Commission (NRC) HTGR research initiative, a need was identified for validation of systems-level computer code modeling capabilities in anticipation of the eventual need to perform licensing analyses. Because the NRC has used MELCOR for light water reactors (LWR) in the past and because MELCOR was recently updated to include gas-cooled reactor (GCR) physics models, MELCOR is among the system codes of interest to the NRC. This paper describes MELCOR modeling of the General Atomics' Modular High Temperature Gas-Cooled Reactor (MHTGR). The MHGTR is a suitable design for demonstration of MELCOR GCR modeling competency for two reasons: 1) the MHTGR is a predecessor to the more advanced General Atomics’ Gas-Turbine Modular High Temperature Reactor (GTMHR), and 2) experimental data useful for benchmark calculations may soon become available. Using the most complete literature references available for the MHTGR design, researchers at Texas A&M University (TAMU) constructed a MELCOR input deck for the MHTGR to partially validate MELCOR GCR modeling capabilities. Normal and off-normal system operating conditions were modeled with appropriate boundary and initial conditions. MELCOR predictions of system response were obtained for steady-state, pressurized conduction cool-down (PCC), and depressurized conduction cool-down (DCC) scenarios. Code results were checked against nominal MHTGR design parameters, physical intuition, and anticipated GCR thermal hydraulic response. No inherent deficiencies in MELCOR modeling capability were observed, suggesting that the newly-implemented GCR models are adequate for systems-level analysis. If and when experimental benchmark data becomes available, further validation activities may proceed given the modeling efforts discussed herein.     

Potential protein degradation balance and total metabolizable protein supply to dairy cows from heat‐treated faba beans

The effects of pressure toasting (100, 118 and 136 °C for 3, 7, 15 and 30 min) on potential protein nutritional value of faba beans were evaluated with the NRC 2001 dairy model, by determining undegraded (RUP) and degraded rumen protein (RDP), undegraded (RUST) and degraded rumen starch (RDST), truly absorbed undegraded protein (ARUP), microbial protein (MCP_RDP) synthesized in the rumen from rumen‐available protein, truly absorbed rumen synthesized microbial protein (AMCP), truly absorbed rumen endogenous protein (AECP), total metabolizable protein (MP) in the small intestine, and the protein degradation balance (PDB). The treatments increased RUP, RUST, ARUP and MP (p < 0.001), and decreased RDP, RDST, MCP_RDP and PDB (p < 0.001), the effects increasing with increasing temperature and time. The treatments increased (p < 0.001) ARUP without affecting AECP and AMCP, so that the net absorbable total MP in the small intestine was increased. The PDB was reduced (p < 0.001) but never became negative. These results indicated that potential microbial protein synthesis would not be impaired due to sufficient nitrogen in the rumen, but the high positive PDB values with most treatments, except 136 °C for 15 min (PDB 2.0 g kg^?1 DM) indicated that there were large potential losses of nitrogen in the rumen, particularly for the control with a value of 88.9 g kg^?1 dry matter. It is concluded that predicted potential protein degradation balance and total metabolizable protein supply from faba beans were improved by the treatments. Copyright © 2005 Society of Chemical Industry     

Analysis of different containment models for IRIS small break LOCA, using GOTHIC and RELAP5 codes

Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones.In this work the international reactor innovative and secure (IRIS) was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a small break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. A simplified model of IRIS containment drywell was implemented with RELAP5 according to a sliced approach, based on the two-pipe-with-junction concept, while it was addressed with GOTHIC using several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, heat transfer coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flow rate. The objective of the paper is to preliminarily compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix (comprising SBLOCA and to some extent LBLOCA transients, but not explicitly the modelling of large dry containment volumes).The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due, e.g. to a marked overestimation of internal natural recirculation, RELAP5 confirmed its capability to satisfactorily model the basic processes in IRIS containment following SBLOCA.     

A review on the clad failure studies

In this paper, an attempt has been made to systematically organize the research investigations conducted on clad tube failure, so far. Before presenting the review on the clad failure studies, an introduction to different clad materials has been added, in which the effect of alloying elements on the material properties have been presented. The literature on clad failure has been broadly categorized under the headings LOCA and RIA. The failure mechanisms like creep, corrosion and pellet-clad interaction have been discussed in details. Each subsection of the review has been provided with summary table, in which the studies are arranged in the chronological order. A small section on acceptance criteria for ECCS has also been included. The last section of the review has been dedicated to the core-degradation phenomena.     

Integrated software safety analysis method for digital I&C systems

The digitalized Instrumentation and Control (I&C) system of Nuclear power plants can provide more powerful overall operation capability, and user friendly man-machine interface. The operator can obtain more information through digital I&C system. However, while I&C system being digitalized, three issues are encountered: (1) software common-cause failure, (2) the interaction failure between operator and digital instrumentation and control system interface, and (3) the non-detectability of software failure. These failures might defeat defense echelons, and make the Diversity and Defense-in-Depth (D3) analysis be more difficult. This work developed an integrated methodology to evaluate nuclear power plant safety effect by interactions between operator and digital I&C system, and then propose improvement recommendations. This integrated methodology includes component-level software fault tree, system-level sequence-tree method and nuclear power plant computer simulation analysis. Software fault tree can clarify the software failure structure in digital I&C systems. Sequence-tree method can identify the interaction process and relationship among operator and I&C systems in each D3 echelon in a design basis event. Nuclear power plant computer simulation analysis method can further analyze the available backup facilities and allowable manual action duration for the operator when the digital I&C fail to function. Applying this methodology to evaluate the performance of digital nuclear power plant D3 design, could promote the nuclear power plant operation safety. The operator can then trust the nuclear power plant than before, when operating the highly automatic digital I&C facilities.     

The application of nonlinear programming on ration formulation for dairy cattle

The objective of this study was to compare the application of iterative linear programming (iteLP), sequential quadratic programming (SQP), and mixed-integer nonlinear programming-based deterministic global optimization (MINLP_DGO) on ration formulation for dairy cattle based on Nutrient Requirements of Dairy Cattle (NRC, 2001). Least-cost diets were formulated for lactating cows, dry cows, and heifers. Nutrient requirements including energy, protein, and minerals, along with other limitations on dry matter intake, neutral detergent fiber, and fat were considered as constraints. Five hundred simulations were conducted, with each simulation randomly selecting 3 roughages and 5 concentrates from the feed table in NRC (2001) as the feed resource for each of 3 animal groups. Among the 500 simulations for lactating cows, 57, 45, and 21 simulations did not yield a feasible solution when using iteLP, SQP, and MINLP_DGO, respectively. All the simulations for dry cows and heifers were feasible when using SQP and MINLP_DGO, but 49 and 11 infeasible simulations occurred when using iteLP for dry cows and heifers, respectively. The average ration costs per animal per day of the feasible solutions obtained by iteLP, SQP, and MINLP_DGO were $4.78 (±0.71), $4.45 (±0.65), and $4.44 (±0.65) for lactating cows; $2.39 (±0.52), $1.48 (±0.26), and $1.48 (±0.26) for dry cows; and $0.98 (±0.72), $0.97 (±0.15), and $0.91 (±0.14) for heifers, respectively. The average computation time of iteLP, SQP, and MINLP_DGO were 0.59 (±1.87) s, 1.15 (±0.62) s, and 58.69 (±68.45) s for lactating cows; 0.041 (±0.070) s, 0.76 (±0.37) s, and 14.84 (±39.09) s for dry cows; and 1.60 (±2.90) s, 0.51 (±0.19) s, and 16.45 (±45.56) s for heifers, respectively. In conclusion, iteLP had limited capability of formulating least-cost diets when nonlinearity existed in the constraints. Both SQP and MINLP_DGO handled the nonlinear constraints well, with SQP being faster, whereas MINLP_DGO was able to return a feasible solution under some situations where SQP could not.     

Activities for 4S USNRC licensing

4S (Super-Safe, Small and Simple) is a small sized sodium-cooled fast reactor being developed for the electricity supply in remote areas, high-temperature steam supply more than 400 °C, seawater desalination, and hydrogen production. The system design of power output of 10 MWe (30 MWt) has been completed. The main feature is that it does not have to be refueled for a long period (i.e. 30 years for 10 MWe version), and enable the reactor closure sealed during plant operation. Furthermore, the small size of the reactor makes the reactor building suitable for below grade installing. These two features can provide resolutions for the issues relevant to safety, security, and safeguard, which become much more serious matter internationally these days.4S is a pool-type reactor which contains the whole primary cooling system in a vessel. For the purpose of reducing the maintenance requirements with the reactor, (1) reflectors to compensate for fuel burn-up instead of control rods, (2) electromagnetic pump (EMP) which has no rotating parts, and (3) residual heat removal system by natural circulation and natural air draft are adopted. Therefore, exchange of the reactor components is not required during plant operation, in addition to no needs for refueling.Toshiba has initiated the U.S. Nuclear Regulatory Commission (NRC) pre-application review of 10 MWe version for the purpose of applying for design approval (DA). A series of public meetings with NRC has been held four times, and five technical reports have been submitted to NRC in preparation for DA application. Topics discussed in these meetings included, plant design, metallic fuel, safety design philosophy, safety analysis, measures against severe accident, phenomena identification and ranking table (PIRT), etc. Some useful comments and questions on the issues regarding the specific feature of 4S as well as sodium-cooled fast reactor were raised by NRC at the public meetings. Among them, those items which are applicable to general sodium-cooled fast reactors, e.g. principal design criteria, guideline for safety analysis, validation and verification for safety analysis code, quality requirements, severe accident, and emergency planning are presented in this paper.     

Grafting of polymer chains on the surface of carbon nanotubes via nitroxide radical coupling reaction

Poly(butylene succinate) (PBS) was grafted on the surface of TEMPO (2,2,6,6‐tetramethyl‐1‐piperidinyloxy) modified multi‐walled carbon nanotubes (MWCNTs) via a nitroxide radical coupling reaction. TEMPO functionalized MWCNTs (MWCNTs‐g‐TEMPO) were synthesized using the Cu(I)‐catalyzed azide/alkyne click chemistry approach and the covalent bond of the nitroxide moieties onto the MWCNTs was confirmed via electron paramagnetic resonance (EPR) spectroscopy. The PBS grafting on the sidewalls of MWCNTs was carried out in solution via peroxide‐induced formation of macroradicals and it was confirmed by EPR and attenuated total reflectance Fourier transform infrared analysis. Preliminary rheological and calorimetric analyses revealed that the grafting improves both the quality of stress transfer across the polymer ? nanotube interface and the degree of dispersion of the filler, which also exhibited a moderate nucleating action on the PBS. Overall, our results demonstrate that nitroxide radical coupling is an efficient and feasible ‘grafting to’ method to covalently bond polymer chains on MWCNTs with possible advantages in the final properties of the polymer nanocomposites. © 2015 Society of Chemical Industry     

对美国核管会运行经验反馈的研究

研究了美国核管会的运行经验反馈体系和流程,以及与经验反馈工作相关的活动,其中一些成熟做法和经验对我国运行经验反馈工作将起到很好的借鉴作用.