scholarly journals KAJIAN KESELAMATAN PENGOPERASIAN REAKTOR TRIGA 2000 BANDUNG DENGAN MENGGUNAKAN BATANG KENDALI REAKTOR TRIGA 2000 TANPA BAHAN BAKAR (BKRTTBB)

2015 ◽  
Vol 16 (2) ◽  
pp. 93
Author(s):  
Prasetyo Basuki ◽  
A.R. Iso Suwarso ◽  
Agus Sunarya ◽  
P. Ilham Yazid ◽  
Mr. Supardjo ◽  
...  
Keyword(s):  
Water Temperature ◽  
Power Factor ◽  
Peak Power ◽  
Reactor Operation ◽  
Primary Coolant ◽  
Control Rod ◽  
Channel Temperature ◽  
Temperature Exchange ◽  
Post Modification ◽  
Fuel Elements

ABSTRAKKAJIAN KESELAMATAN PENGOPERASIAN REAKTOR TRIGA 2000 BANDUNG DENGAN MENGGUNAKAN BATANG KENDALI REAKTOR TRIGA 2000 TANPA BAHAN BAKAR (BKRTTBB). Telah dilakukan kegiatan pabrikasi BKRTTBB untuk digunakan pada teras TRIGA 2000 Bandung sebagai upaya modifikasi batang kendali pengganti FFCR (Fuel Follower Control Rod) yang telah memiliki fraksi bakar melebihi 50% pada bagian elemen pengikutnya. Dari 5 buah FFCR yang digunakan saat ini telah terindikasi 2 buah yang memiliki fraksi bakar melebihi 50 % dan 1 buah yang telah mendekati 50 %. Sampai dengan akhir tahun ini direncanakan dilakukan penggantian sebanyak 2 buah, dan akan berlanjut sampai dengan 4 buah di tahun berikutnya. Untuk dapat menjamin keselamatan proses modifikasi dan pasca modifikasi, maka perlu dilakukan kajian simulasi operasi dengan menggunakan BKRTTBB pada skenario teras paling reaktif. Pada kajian ini telah dilakukan simulasi operasi dengan meng-gunakan 1 buah FFCR, 4 buah BKRTTBB, dan 102 elemen bakar dengan komposisi elemen bakar sesuai dengan kondisi terkini pada teras TRIGA 2000 dengan menggunakan MCNP. Dari kajian ini didapatkan beberapa parameter kritikalitas antara lain reaktivitas teras lebih (core excess) sebesar $ 5,461, dan reaktivitas padam (shutdown margin) sebesar $ -9,647, kemudian dengan menskenariokan kondisi one stuck rod didapatkan bahwa semua kondisi salah satu batang kendali tersangkut memberikan respons subkritis. Kemudian dari simulasi ini pula di-dapatkan faktor puncak daya aksial sebesar 1,21 dan faktor puncak daya radial sebesar 2,02. Dari kedua nilai faktor puncak daya ini dapat dihitung distribusi suhu pada teras dengan menggunakan program komputasi STAT dan STATMOD. Hasil simulasi menggunakan STAT dan STATMOD dengan memberikan suhu masukan air sampai dengan 42 °C didapatkan suhu terpanas pada subbuluh sebesar 87,98 °C dan 82,75 °C. Berdasarkan hasil ini dapat disimpul-kan bahwa pengoperasian reaktor dengan menggunakan BKRTTBB pada kondisi yang men-dekati dimana suhu air masukan mendekati 49 °C (suhu tertinggi untuk sinyal SCRAM), air pendingin primer belum mencapai suhu pendidihan (112 °C). Sehingga pengoperasian reaktor dengan BKRTTBB dapat dinyatakan aman dan selamat dari aspek neutronik maupun termal-hidrolik.ABSTRACTSAFETY REVIEW OF BANDUNG TRIGA 2000 RESEARCH REACTOR OPERATION USING CONTROL ROD WITHOUT FUEL FOLLOWER (BKRTTBB). BKRTTBB manufacturing activities have been carried out to be used on the TRIGA 2000 core as a modification of the control rod replacement FFCR (Fuel Follower Control Rod) which has had burnup exceeds 50 % on the fuel follower. Two units of existing FFCR have been indicated exceeds 50 % of burnup and 1 unit was approaching 50%. Until the end of this year planned replacement by 2 units, and will continue up to 4 units in the next year. To ensure the safety of the modification process and the post-modification activities, it is necessary to study the operation simulation using BKRTTBB on the most reactive core. This study has been carried out on simulated reactor operation using 1 unit FFCR, 4 units BKRTTBB, and 102 fuel elements with composition in accordance with current conditions on the TRIGA 2000 core by using MCNP. This study obtained some criticality parameters, core excess $ 5.461, and shutdown margin $ -9.647, then the scenario of one stuck rod conditions showed that all the conditions of one control rod stuck is responded as subcritical. Then from this simulation also obtained axial peak power factor of 1.21 and radial peak power factor of 2.02. Based on these values, the temperature distribution on the reactor can be calculated using computational codes, STAT and STATMOD. The simulation results using STAT and STATMOD by providing input water temperature up to 42 °C at the hottest sub channel temperature obtained of 87.98 °C and 82.75 °C. Based on these results it can be concluded that operation of the reactor by using BKRTTBB in conditions near to LOFA where no temperature exchange so that the water temperature input approaching 49 °C (the highest temperature for the scram signal), water primary coolant still has not reached the temperature of boiling (112 °C). So that the operation of the reactor with BKRTTBB can be declared safely and secure in neutronics and termalhydraulics aspect.

Root Cause Analysis of SI Nozzle Thermal Sleeve Breakaway Failures Occurring at PWR Plants

2008 ◽  
Vol 131 (1) ◽  
Author(s):  
Jong Chull Jo ◽  
Myung Jo Jhung ◽  
Seon Oh Yu ◽  
Hho Jung Kim ◽  
Young Gill Yune
Keyword(s):  
Cold Water ◽  
Operation Mode ◽  
Branch Pipe ◽  
Thin Wall ◽  
Reactor Operation ◽  
Support Condition ◽  
Primary Coolant ◽  
Root Cause ◽  
Main Pipe ◽  
Flow Situation

At conventional pressurized water reactors (PWRs), cold water stored in the refueling water tank of emergency core cooling system is injected into the primary coolant system through a safety injection (SI) line, which is connected to each cold leg pipe between the main coolant pump and the reactor vessel during the SI operation, which begins on the receipt of a loss of coolant accident signal. In normal reactor power operation mode, the wall of SI line nozzle maintains at high temperature because it is the junction part connected to the cold leg pipe through which the hot main coolant flows. To prevent and relieve excessive transient thermal stress in the nozzle wall, which may be caused by the direct contact of cold water in the SI operation mode, a thermal sleeve in the shape of thin wall cylinder is set in the nozzle part of each SI line. Recently, mechanical failures that the sleeves are separated from the SI branch pipe and fall into the connected cold leg main pipe occurred in sequence at some typical PWR plants in Korea. To find out the root cause of thermal sleeve breakaway failures, the flow situation in the junction of primary coolant main pipe-SI branch pipe and the vibration modal characteristics of the thermal sleeve are investigated in detail by using both computational fluid dynamics code and structure analysis finite element code. As a result, the transient response in fluid pressure exerting on the local part of thermal sleeve wall surface to the primary coolant flow through the pipe junction area during the normal reactor operation mode shows oscillatory characteristics with the frequencies ranging from 15Hzto18Hz. These frequencies coincide with the lower mode natural frequencies of thermal sleeve, which has a pinned support condition on the outer surface with the circumferential prominence set into the circumferential groove on the inner surface of SI nozzle at the midheight of thermal sleeve. In addition, the variation of pressure on the thermal sleeve surface yields alternating forces and torques in the directions of two rectangular axes perpendicular to the longitudinal axis of cylindrical thermal sleeve, which causes both rolling and pitching motions of the thermal sleeve. Consequently, it is seen that this flow situation surrounding the thermal sleeve during the normal reactor operation can induce resonant vibrations accompanying the shaking motion of the thermal sleeve at the pinned support condition, which finally leads to the failures of thermal sleeve breakaway from the SI nozzle.

scholarly journals Study of Hydrogen Generation of Aluminum-Containing Compositions with Boric Acid

2018 ◽  
Vol 3 (3) ◽  
pp. 230
Author(s):  
Salakhova A.A. ◽  
Suvorov V.A. ◽  
Firsova A. I. ◽  
Belozerov V.I. ◽  
Milinchuk V.K.
Keyword(s):  
Boric Acid ◽  
Hydrogen Generation ◽  
Sodium Metasilicate ◽  
Chemical Processes ◽  
System Management ◽  
Control Rod ◽  
Radiation Fields ◽  
Fuel Elements ◽  
And Control ◽  
Kinetics Of

The results of investigations of the kinetics of hydrogen generation compositions with aluminum, chemical activators (hydrated sodium metasilicate, oxide and calcium hydroxide) boric acid. Aluminium and its alloys used for the manufacture of protective sheaths of fuel elements and control rod protection system management, pipelines, tanks, and various support structures in the active zone of atomic reactors RBMK, research water-cooled reactors. The aluminum is protected from direct contact with water and steam surface layer of metal oxide having a high corrosion resistance at high temperatures in powerful radiation fields. However, after removal or when the discontinuity of the oxide layer of activated metal efficiently decompose water to hydrogen. It is established that the hydrogen aluminum-containing compositions is dependent on the concentration of boric acid. The discovery of the involvement of boric acid in these reactions expands the ideas about regularities of chemical processes of formation of hydrogen flowing in the water coolant of VVER reactors with the participation of the corrective additives and impurities.

On Efficient Passive Cooling of Control Rod Drivers

2013 ◽  
Author(s):  
A. N. Gershuni ◽  
A. P. Nishchik ◽  
V. G. Razumovskiy ◽  
I. L. Pioro
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
High Performance ◽  
Cooling System ◽  
Vertical Channel ◽  
Passive Cooling ◽  
Primary Circuit ◽  
Two Phase ◽  
Primary Coolant ◽  
Control Rod

Experimental research of natural convection and the ways of its suppression in an annular vertical channel to simulate the conditions of cooling the control rod drivers of the reactor protection system (RPS) in its so-called wet design, where the drivers are cooled by primary circuit water supplied due to the system that includes branched pipelines, valves, pump, heat exchanger, etc., is reported. Reliability of the drivers depends upon their temperature ensured by operation of an active multi-element cooling system. Its replacement by an available passive cooling system is possible only under significant suppression of natural convection in control rod channel filled with primary coolant. The methods of suppression of natural convection proposed in the work have demonstrated the possibility both of minimization of axial heat transfer and of almost complete elimination of temperature non-uniformity and oscillation inside the channel under the conditions of free travel of moving element (control rod) in it. The obtained results widen the possibilities of substitution of the active systems of cooling the RPS drivers by reliable passive systems, such as high-performance heat-transfer systems of evaporation-condensation type with heat pipes or two-phase thermosyphons as heat-transferring elements.

scholarly journals Achieving high power factor and output power density in p-type half-Heuslers Nb1-xTixFeSb

2016 ◽  
Vol 113 (48) ◽  
pp. 13576-13581 ◽  
Author(s):  
Ran He ◽  
Daniel Kraemer ◽  
Jun Mao ◽  
Lingping Zeng ◽  
Qing Jie ◽  
...  
Keyword(s):  
Power Density ◽  
Output Power ◽  
Power Factor ◽  
High Power ◽  
Peak Power ◽  
Large Scale ◽  
The Past ◽  
High Power Factor ◽  
P Type ◽  
High Output

Improvements in thermoelectric material performance over the past two decades have largely been based on decreasing the phonon thermal conductivity. Enhancing the power factor has been less successful in comparison. In this work, a peak power factor of ∼106 μW⋅cm−1⋅K−2is achieved by increasing the hot pressing temperature up to 1,373 K in the p-type half-Heusler Nb0.95Ti0.05FeSb. The high power factor subsequently yields a record output power density of ∼22 W⋅cm−2based on a single-leg device operating at between 293 K and 868 K. Such a high-output power density can be beneficial for large-scale power generation applications.

Root Cause Analysis of SI Nozzle Thermal Sleeve Breakaway Failures Occurred at PWR Plants

2006 ◽  
Author(s):  
Jong Chull Jo ◽  
Myung Jo Jhung ◽  
Seon Oh Yu ◽  
Hho Jung Kim ◽  
Young Gill Yune
Keyword(s):  
Branch Pipe ◽  
Hot Water ◽  
Thin Wall ◽  
Reactor Operation ◽  
Support Condition ◽  
Primary Coolant ◽  
Lower Mode ◽  
Root Cause ◽  
Main Pipe ◽  
Flow Situation

Thermal sleeves in the shape of thin wall cylinder seated inside the nozzle part of each safety injection (SI) line at pressurized water reactors (PWRs) have such functions as prevention and relief of potential excessive transient thermal stress in the wall of SI line nozzle part which is initially heated up with hot water flowing in the primary coolant piping system when cold water is injected into the system through the SI nozzles during the SI operation. Recently, mechanical failures that the sleeves are separated from the SI branch pipe and fall into the connected cold leg main pipe occurred in sequence at some typical PWR plants in Korea. To find out the root cause of thermal sleeve breakaway failures, the flow situation in the in the junction of primary coolant main pipe and SI branch pipe and the vibration modal characteristics of the thermal sleeve are investigated in details by using both computational fluid dynamic (CFD) code and structure analysis finite element code. As the results, the transient response in fluid pressure exerting on the local part of thermal sleeve wall surface to the primary coolant flow through the pipe junction area during the normal reactor operation mode shows oscillatory characteristics with the frequencies ranging from 15 to 18, which coincide with the lower mode natural frequencies of thermal sleeve having a pinned support condition on the circumferential prominence on the outer surface of thermal sleeve which is put into the circumferential groove on the inner surface of SI nozzle at the mid-height of thermal sleeve. In addition, the variation of pressure on the thermal sleeve surface yield alternating forces and torques in the directions of two rectangular axes perpendicular to the longitudinal axis of cylindrical thermal sleeve, which causes both rolling and pitching motions of the thermal sleeve. Consequently, it is seen that this flow situation surrounding the thermal sleeve during the normal reactor operation can induce resonant vibrations accompanying the shaking motion of the thermal sleeve at the pinned support condition, which finally leads to the failures of thermal sleeve breakaway from the SI nozzle.

Neutronic Design Features of a Transportable Fluoride-Salt-Cooled High-Temperature Reactor

2016 ◽  
Vol 2 (3) ◽  
Author(s):  
Kaichao Sun ◽  
Lin-Wen Hu ◽  
Charles Forsberg
Keyword(s):  
High Temperature ◽  
Peak Power ◽  
Fuel Cycle ◽  
Thermal Power ◽  
Combined Cycle ◽  
Fluoride Salt ◽  
Control Rod ◽  
The Core ◽  
Liquid Salt ◽  
High Temperature Reactor

The fluoride-salt-cooled high-temperature reactor (FHR) is a new reactor concept, which combines low-pressure liquid salt coolant and high-temperature tristructural isotropic (TRISO) particle fuel. The refractory TRISO particle coating system and the dispersion in graphite matrix enhance safeguards (nuclear proliferation resistance) and security. Compared to the conventional high-temperature reactor (HTR) cooled by helium gas, the liquid salt system features significantly lower pressure, larger volumetric heat capacity, and higher thermal conductivity. The salt coolant enables coupling to a nuclear air-Brayton combined cycle (NACC) that provides base-load and peak-power capabilities. Added peak power is produced using jet fuel or locally produced hydrogen. The FHR is, therefore, considered as an ideal candidate for the transportable reactor concept to provide power to remote sites. In this context, a 20-MW (thermal power) compact core aiming at an 18-month once-through fuel cycle is currently under design at Massachusetts Institute of Technology (MIT). One of the key challenges of the core design is to minimize the reactivity swing induced by fuel depletion, since excessive reactivity will increase the complexity in control rod design and also result in criticality risk during the transportation process. In this study, burnable poison particles (BPPs) made of B4C with natural boron (i.e., 20% B10 content) are adopted as the key measure for fuel cycle optimization. It was found that the overall inventory and the individual size of BPPs are the two most important parameters that determine the evolution path of the multiplication factor over time. The packing fraction (PF) in the fuel compact and the height of active zone are of secondary importance. The neutronic effect of Li6 depletion was also quantified. The 18-month once-through fuel cycle is optimized, and the depletion reactivity swing is reduced to 1 beta. The reactivity control system, which consists of six control rods and 12 safety rods, has been implemented in the proposed FHR core configuration. It fully satisfies the design goal of limiting the maximum reactivity worth for single control rod ejection within 0.8 beta and ensuring shutdown margin with the most valuable safety rod fully withdrawn. The core power distribution including the control rod’s effect is also demonstrated in this paper.

Transient Analysis of UO2 and MOX-Fueled Cores Based on Microscopic Reactor Physics

2008 ◽  
Author(s):  
Toshikazu Takeda ◽  
Hiroaki Tagawa ◽  
Tadafumi Sano
Keyword(s):  
Cross Sections ◽  
Present Method ◽  
Peak Power ◽  
Transient Analysis ◽  
Effective Cross Section ◽  
Light Water Reactor ◽  
Reactor Physics ◽  
Control Rod ◽  
Fuel Rod ◽  
Averaged Model

A transient analysis has been performed for UO2 and MOX-fueled light water reactor cores based on Microscopic Reactor Physics, which treats the detailed distribution of temperature and effective cross section within a rod. Conventionally the volume-averaged temperature and the Rowlands’ effective temperature are used to calculate fuel rod-averaged cross sections, and applied to the transient analysis. The present method is considered as a reference and the result is compared with the conventional method for a mini fuel core containing eight fuel rods and a control rod. From numerical results, it is found that the Rowlands’ model underestimates the peak power and the volume averaged model produces rather good peak power results. After 1.0 sec, the Rowlands’ model yields the similar power as the reference, while the volume averaged model yields less power than the reference one.

Experimental Tests With Non-Uniformly Heated 19-Pins Fuel Bundle Cooled by HLM

2018 ◽  
Author(s):  
M. Angelucci ◽  
I. Di Piazza ◽  
M. Tarantino ◽  
R. Marinari ◽  
G. Polazzi ◽  
...  
Keyword(s):  
Power Distribution ◽  
Natural Circulation ◽  
Experimental Tests ◽  
Primary Coolant ◽  
Experimental Campaign ◽  
Fuel Bundle ◽  
Channel Temperature ◽  
Fuel Pin ◽  
System Data ◽  
Local Temperature Distribution

An experimental campaign was performed on a non-uniformly heated 19-pins wire-spaced fuel pin bundle simulator, cooled by Heavy Liquid Metal and installed in the NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy). The experimental tests concerned mass flow rate transition of the primary coolant from forced to natural circulation, with fuel pin bundle simulator characterized by non-uniform power distribution. The main objective of the performed experimental campaign was to perform integral system and local thermal-hydraulic analysis, in particular to investigate the flow in different flow regimes and specifically the transition from forced to natural circulation flow and, more specifically, analyze the behavior of the 19-pins wire-spaced fuel pin simulator (FPS) during such transient. Indeed, the performed test were characterized by non-uniform heating of the bundle (i.e. just some pins switched on), so the effects of this non-uniformity on the local temperatures and on the overall system behavior was evaluated. A deep investigation on the local temperature distribution was performed thanks to the accurate instrumentation provided in the bundle (67 thermocouples). For instance, in some cases, the wall temperatures relative to pins switched off remained below the relative sub-channel temperature, depending on the heating distribution. The obtained experimental data provided useful information for the characterization of the bundle and the computation of the heat transfer coefficient. Moreover, the collected system data can be helpful for STH codes validation, whereas the local fuel bundle data, especially the ones from dissymmetric tests can be useful for the qualification and benchmarking of CFD codes and coupled STH/CFD methods for HLM systems.

Buildup and Decay Analysis of Corrosion Products Activity in Primary Coolant Loop of AP-1000

2018 ◽  
Author(s):  
Fiaz Mahmood ◽  
Huasi Hu ◽  
Liangzhi Cao
Keyword(s):  
Radiation Exposure ◽  
Operating Conditions ◽  
Corrosion Products ◽  
Normal Operation ◽  
Reactor Operation ◽  
Primary Coolant ◽  
Reactor Control ◽  
Power Regulation ◽  
Coolant Loop ◽  
Control Regime

The broad half-life range of Activated Corrosion Products (ACPs) results in major radiation exposure throughout reactor operation and shutdown. The movement of unpredicted activity hot spots in coolant loop can bring about huge financial and dosimetric impacts. The PWR operating experience depicts that activity released during reactor operation and shutdown cannot be estimated through a simple correlation. This paper seeks to analyze buildup and decay behavior of ACPs in primary coolant loop of AP-1000 under normal operation, power regulation and shutdown modes. The application of a well-tested mathematical model is extended in an in-house developed code CPA-AP1000, to simulate the behavior of dominant Corrosion Products (CPs), by programing in MATLAB. The MCNP code is used as a subroutine of the program to model the reactor core and execute energy dependent neutron flux calculations. It is observed that short-lived CPs (56Mn, 24Na) build up rapidly under normal operation mode and decay quickly after the reactor is shutdown. The long-lived CPs (59Fe, 60Co, 99Mo) have exhibited slow buildup under normal operating conditions and likewise sluggish decay after the shutdown. To analyze activity response during reactor control regime, operating power level is promptly decreased and in response specific activity of CPs also followed decreasing trend. It is noticed that activity of CPs drops slowly during reactor control regime in comparison to emergency scram. The results are helpful in estimating radiation exposure caused by ACPs during accessibility of the equipment in coolant loop, under normal operation, power regulation and shutdown modes. Moreover, current analyses provide baseline data for further investigations on ACPs in AP-1000, being a new reactor design.

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