05 Nuclear fuels (scientific, technical)

06•00624 New dynamic method to measure rod worths in zero power physics test at PWR startup Lee, E. K. et al. Annals of Nuclear Energy, 2005, 32, (13), 1457 1475. To measure and val idate the worth of control (or shutdown) bank in zero power physics test at PWRs, a dynamic control rod reactivity measu remen t (DCRM) technique has been developed and appl ied to six s tar t-ups of Wes t inghouse plants as well as Korea Standard Nuclear power Plants based on the Combust ion Eng inee r ing System 80 NSSS. Wi th this technique, jus t one test bank is inser ted into the bo t tom of the core at maximum stepping rate and withdrawn immedia te ly to the all rod-out posit ion. Specially des igned inverse point kinet ics equat ions are used to de te rmine the test bank worth from the measured ex-core detector signals, which are control led by the neut ron- to- response conversion factor and the dynamic-to-stat ic conversion factor. These two paramete r s are p rede te rmined by the th ree-d imens iona l neut ron adjoint flux dis t r ibut ion for both the top and bot tom ex-core detector and the th ree-d imens iona l s teady and t rans ient core power dis t r ibut ion for tes t bank movement . To e l iminate the gamma-ray effect on ex-core detector signals, a s imple method, using reactivity curve characterist ics, was also developed. To verify the D C R M method, a total of 28 bank worths of six different P W R s was measured by the D C R M and compared with those of convent ional method. Resul ts show that the D C R M method has a s imilar accuracy as the convent ional technique. However, with the D C R M method, it only takes a r o u n d l 5 min per bank from the beginning of rod inser t ion to the de te rmina t ion of measured static worth. F rom its performance, one can conclude that the D C R M method is an effective rep lacement for the convent ional rod worth measuremen t method.

06•00625 Nuclear power for sustainable development and relevant IAEA activities for the future Omoto, A. Progress in Nuclear Energy, 2005, 47, (1 4), 16 26. The credible longer- te rm energy demand and supply analyses foresee a growing role for nuclear power for sustainable development . For instance the Special Repo r t on Emiss ions Scenarios of the Inter- governmenta l Panel on Cl imate Change ( IPCC) shows an increase between 2000 and 2050 by a factor of 2.5 in global pr imary energy and the instal led nuclear capacity will increase by about a factor of 4 5 as a median value. The technologies for the nuclear energy are cont inuously improving towards the long- term goals of fur ther improvements in economics, very high levels of safety, increased prol i ferat ion resistance, and successful implementa t ion of solut ions for radioact ive waste disposal. By statute, the I A E A is author ized to encourage and assist the Member States efforts for the practical appl icat ion of nuclear technology. The Agency 's re levant activities are considered to contr ibute to assist the Member States to achieve their long- term goals. This paper overviews the current s tatus of nuclear power in the world, discusses its future prospects and describes the I A E A ' s activities to suppor t its Member States in their efforts for nuclear p rogramme for susta inable development .

06•00626 OECD/NEA activities relating to innovative nuclear energy systems Marcus, G. H. Progress in Nuclear Energy, 2005, 47, (1 4), 27 31. The mission of the O E C D Nuclear Energy Agency (NEA) is to assist its member countr ies in main ta in ing and further developing, through in te rna t iona l co-operat ion, the scientific, technological and legal bases requi red for the safe, envi ronmenta l ly fr iendly and economical use of nuclear energy for peaceful purposes. In fulfilling tha t mission, the N E A conducts technical, economic and policy studies in response to the needs and interests of its 28 member countries. In recent years, a number of these studies have addressed var ious aspects of the next genera t ion of nuclear power plants. This paper will describe some of the major activities recently comple ted and current ly underway that may be of par t icular interest to the C O E - I N E S program.

06•00627 On the possibility of the space-dependence of the stability indicator (decay ratio) of a BWR Demazi~re, C. et al. Pitzsit, I. Annals of Nuclear Energy, 2005, 32, (12), 1305 1322. A model is proposed for the explanat ion of the space-dependence of the so-called decay rat io (DR) which is used to quantify the stabil i ty proper t ies of boil ing water reactors (BWRs). The study was p rompted by the observat ion of a strongly space-dependent decay rat io in an instabil i ty event at the Swedish Forsmark-1 BWR. Prior to that event, the space-dependence of the D R was ne i ther observed, nor assumed possible in the theoret ical models of instability. The model proposed here is based on a previous suggest ion by one of the authors on how to model the es t imat ion of the DR in case of two different types of osci l lat ions (instabil i t ies) be ing present in the core s imultaneously. The model was earl ier only used in a space- independent form, but here its applicabi l i ty is extended such that space-dependence of the oscil lat ions is also accounted for, by using a noise s imulator . The invest igat ions show tha t the DR, as de te rmined by the individual LPRMs (neut ron detectors) at different positions, can be strongly space-dependent if at

96 Fuel and Energy Abstracts March 2006

least two different osci l lat ions with differing D R and space-depen- dence exist in the core simultaneously. The observed space-dependence of the D R in the Forsmark case can be reconst ructed by the model.

06•00628 Parametric studies on different gas turbine cycles for a high temperature gas-cooled reactor Wang, J. and Gu, Y. Nuclear Engineering ancl Design, 2005, 235, (16), 1761 1772. The high t empera tu re gas-cooled reactor ( H T G R ) coupled with turbine cycle is considered as one of the leading candidates for future nuclear power plants. In this paper, the var ious types of H T G R gas turbine cycles are concluded as three typical cycles of direct cycle, closed indirect cycle and open indirect cycle. Fur the rmore they are theoret ical ly converted to three Brayton cycles of hel ium, ni t rogen and air. Those three types of Brayton cycles are thermodynamical ly analysed and optimized. The results show tha t the var ie ty of gas affects the cycle pressure rat io more significantly than other cycle parameters , however, the opt imized cycle efficiencies of the three Brayton cycles are a lmost the same. In addit ion, the turbo machines that are required for the three opt imized Brayton cycles are aerodynamical ly analysed and compared and their fundamenta l character is t ics are obtained. He l ium turbo-compressor has lower stage pressure ra t io and more stage number than those for ni t rogen and air machines, while he l ium and ni t rogen turbo-compressors have shorter b lade length than that for air machine.

06•00629 Prediction of fission mass-yield distributions based on cross section calculations Hambsch, F.-J. et al. Annals of Nuclear Energy, 2005, 32, (12), 1297 1304. For the first time, fission mass-yield dis t r ibut ions have been predic ted

based on an extended stat ist ical model for fission cross-section calculations. In this model , the concept of the mul t i -modal i ty of the fission process has been incorporated. The three most dominan t fission modes, the two asymmetr ic s tandard I ($1) and s tandard II ($2) modes and the symmetr ic super long (SL) mode are taken into account. De- convoluted fission cross sections for $1, $2 and SL modes for 23s'238U(n, f) and 237Np(n, f), based on exper imenta l branching ratios, were calculated for the first t ime in the incident neu t ron energy range from 0.01 to 5.5 MeV providing good ag reemen t with the exper imenta l fission cross section data. The branching rat ios obta ined from the modal fission cross section calculat ions have been used to deduce the corresponding fission yield distr ibutions, including mean values also for incident neu t ron energies h i ther to not accessible to exper iment .

06•00630 Radiation damage studies on the first wall of a HYLIFE-II type fusion breeder Sahin, S. and Ubeyli , M. Energy Conversion and Management, 2005, 46, (20), 3185 3201. The radia t ion damage on the first wall [made of (1) a ferri t ic s teel (9Cr 2WVTa), (2) a vanad ium alloy ( V 4 C r 4Ti) and (3) S iCjS iC composi te] of an iner t ia l fusion energy ( IFE) reactor of H Y L I F E - I I type is investigated. A protect ive l iquid wall with var iable thickness, containing Flibe + heavy meta l sal t (UF4 or ThF4) is used for first wall protect ion. The content of heavy meta l sal t is chosen as 4 and 12 mol%. Neut ron t ranspor t calculat ions are performed with the aid of the SCALE4.3 System by solving the Bol tzmann t ranspor t equat ion with the X S D R N P M code in 238 energy groups and S~ P3 approximat ion. A flowing wall with a thickness of ~60 cm can extend the l i fet ime of the solid first wall s t ructure to a p lant l i fet ime of 30 years for 9Cr 2WVTa and V 4 C r 4Ti, whereas the S iCjS iC composi te as first wall needs a flowing wall with a thickness of ~85 cm to mainta in the radia t ion damage limit. Substant ia l extra revenue can be gained through the inser t ion of a heavy meta l sal t const i tuent into Flibe, which allows breeding fissile fuel for external reactors and increasing energy

33 mul t ip l icat ion (2 U with a value of up to $1,000,O00,O00/year or 39 2 Pu for few $100,O00,O00/year ).

06•00631 RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320 Gencheva, R. V. et al. Annals of Nuclear Energy, 2005, 32, (12), 1407 1434. Dur ing the deve lopment of Symptom Based Emergency Opera t ing Procedures (SB-EOPs) for VVER-1000/V320 units a t Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is ' Invest igat ion of reactor vessel Y R line capabil i t ies for pr imary side depressur iza t ion dur ing the total loss of feed water (TLFW)' . The main purpose of these calculat ions is to evaluate the capabil i t ies of Y R line located at the top of the reactor vessel for pr imary side depressuriza- t ion to the set point of high pressure injection system (HPIS) actuat ion and the abil i t ies for successful core cooling after Feed & Bleed procedure init iation. For the purpose of this, opera tor act ion with 'Reac tor vessel off-gas valve 0.032 m' opening has been investigated.