Perancangan Pipa Bawah Laut Mobil Cepu Ltd. di Perairan Tuban, Jawa Timur

Tugas Akhir ini merancang dan menganalisa pipa bawah laut Mobil Cepu Ltd. di perairan Tuban, Jawa Timur. Beban-beban hidrodinamis yang mengenai pipa akan mempengaruhi kestabilan dan kekuatan struktur pipa. Perancangan meliputi perhitungan diameter pipa, ketebalan pipa, ketebalan lapisan beton (concrete coating) serta densitas lapisan beton. Analisa pada pipa meliputi analisa buckling dan analisa kekuatan struktur pipa. Perancangan dan ana/isa yang dilakukan berdasarkan pada referensi yang berkaitan dengan perancangan pipa bawah /aut serta rules yang diberikan oleh Germanischer Lloyd (1995) Rules for Classification and Construction, Part 4-Subsea Pipelines and Riser. Daerah studi berada di pantai utara perairan Tuban, Jawa Timur. Tinggi Ge/ombang sinifikan pad a daerah studi adalah 1. 22 meter dengan periode gelombang 5. 73 detik. Dari hasil perhitungan dan analisa dapat disimpulkan bahwa diameter Juar pipa ada/ah 0.457 m (18 in) dengan keteba/an pipa 0.0125 m (0.5 in). Seluruh bagian pipa memiliki ketebalan Japisan beton 7.62 em (3 in) dengan variasi densitas berdasarkan kedalaman. Sistem proteksi trenching/burial direkomendasikan agar pipa tidak menga/ami kerusakan dan ketidakstabilan, khususnya pada kedalaman 1.56 meter sampai 11 meter

Reactor physics design of supercritical CO₂-cooled fast reactors

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.Includes bibliographical references (p. 109-113).Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO₂ (S-CO₂) as a Brayton cycle working fluid in a direct cycle is evaluated. By using S- CO₂ at turbine inlet conditions of 20 MPa and 550⁰C - 700⁰C, efficiencies between 45% and 50% can be achieved with extremely compact components. Neutronic evaluation of candidate core materials was performed for potential use in block-type matrix fueled GFRs with particular concentration on lowering coolant void reactivity to less than $1. SiC cercer fuel was found to have relatively low coolant void worth (+22 cents upon complete depressurization of S-CO₂ coolant) and tolerable reactivity- limited burnup at matrix volume fractions of 60$2.00 in ODS MA956-clad case having H/D of 1).(cont.) It was shown, however, that S-CO₂ coolant void reactivity could be lowered significantly - to less than $1 - in pin cores by increasing neutron leakage (e.g. lowering the core H/D ratio to 0.625 in a pin core with ODS MA956 cladding), an effect not observed in cores using helium coolant at 8 MPa and 500⁰C. An innovative "block"-geometry tube-in-duct fuel consisting of canisters of vibrationally compacted (VIPAC) oxide fuel was introduced and some preliminary calculations were performed. A reference tube-in-duct core was shown to exhibit favorable neutron economy with a conversion ratio (CR) at beginning of life (BOL) of 1.37, but had a coolant void reactivity of +$ 1.4. The high CR should allow designers to lower coolant void worth by increasing leakage while preserving the ability of the core to reach high burnup. Titanium, vanadium and scandium were found to be excellent reflector materials from the standpoint of ... and coolant void reactivity due to their unique elastic scattering cross-section profiles: for example, the SiC cercer core having void reactivity of +$0.22 with titanium was shown to have +$0.57 with Zr₃Si₂.(cont.) Overall, present work confirmed that the S-CO₂-cooled GFR concept has promising characteristics and a sufficiently broad opion space such that a safe and competitive design could be developed in future work with considerably less than $1 void reactivity and a controllable [delta]k due to burnup.by Michael A. Pope.S.M

Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance

Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance