LWR Reactor Research Articles

ICONE15-10617 COMMERICALIZATION OF THE GLOBAL NUCLEAR ENERGY PARTNERSHIP (GNEP)

In February 2006 President Bush announced the Advanced Energy Initiative, which included the Department of Energy's (DOE) Global Nuclear Energy Partnership (GNEP). GNEP has seven broad goals, one of the major elements being to develop and deploy advanced nuclear fuel recycling technology. DOE is contemplating accelerating the deployment of these technologies to achieve the construction of a commercial scale application of these technologies. DOE now defines this approach as "…two simultaneous tracks: (1) deployment of commercial scale facilities for which advanced technologies are available now or in the near future, and (2) further research and development of transmutation fuels technologies." GE believes an integrated technical solution, using existing reactor and fuel reprocessing technologies, is achievable in the near term to accelerate the commercial demonstration of GNEP infrastructure. The concept involves a single, integrated, commercial scale, recycling facility consisting of the Consolidated Fuel Treatment Center (CFTC), capable of processing LWR and fast reactor Spent Nuclear Fuel (SNF) and fabricating Advanced Recycling Reactor (ARR) actinide fuel. The integrated facility would include a fast reactor that uses actinide-bearing fuel to produce electricity. For optimal performance, GE believes this integrated facility should be co-located to eliminate transportation between the CFTC and ARR, and enhance proliferation resistance. This Advanced Recycling Center takes advantage of previous investments by government and industry in fast reactor technology research and development. To allow for commercial acceptance, a prototypical demonstration reactor and associated fuel cycle facility will be constructed, tested, and licensed. Taking advantage of GE's NRC-reviewed modular sodium-cooled PRISM reactor, only a single reactor will be needed and the cost and risk minimized in the initial phase of the program. This paper outlines a process and a schedule to deploy the necessary domestic infrastructure to achieve the goals of GNEP. It is recognized that there are many types of public private partnerships that could demonstrate the technologies to close the fuel cycle. It is important to acknowledge the realities in the marketplace when developing an approach to advance nuclear technologies that can gain widespread commercial acceptance.

Validation of constraint‐based methodology in structural integrity of ferritic steels for nuclear reactor pressure vessels

ABSTRACTVOCALIST (validation of constraint‐based methodology in structural integrity) was a shared cost action project co‐financed by DG Research of the European Commission under the Fifth Framework of the European Atomic Energy Community (EURATOM). The motivation for VOCALIST was based on the understanding that the pattern of crack‐tip stresses and strains causing plastic flow and fracture in components is different to that in test specimens. This gives rise to the so‐called constraint effect. Crack‐tip constraint in components is generally lower than in test specimens. Effective toughness is correspondingly higher. The fracture toughness measured on test specimens is thus likely to underestimate that exhibited by cracks in components. The purpose of VOCALIST was to develop validated models of the constraint effect and associated best practice advice, with the objective of aiding improvements in defect assessment methodology for predicting safety margins and making component lifetime management decisions. The main focus in VOCALIST was an assessment of constraint effects on the cleavage fracture toughness of ferritic steels used in the fabrication of nuclear reactor pressure vessels, because of relevance to the development of improved safety assessments for plant under postulated accident conditions. This paper provides a detailed summary of the main results and conclusions from VOCALIST and points out their contribution to advances in constraint‐based methodology for structural integrity assessment. In particular, the output from VOCALIST has improved confidence in the use of KJ−Tstress and KJ−Q approaches to assessments of cleavage fracture where the effects of in‐plane constraint are dominant. Cleavage fracture models based on the Weibull stress, σW, have been shown to be reliable, although current best practice advice suggests that σW should be computed in terms of hydrostatic stress (as distinct from maximum principal stress) for problems involving out‐of‐plane loading. Correspondingly, the results suggest that the hydrostatic parameter, QH, is the appropriate one with which to characterize crack‐tip constraint in analysing such problems. The materials characterization test results generated as part of VOCALIST have provided added confidence in the use of sub‐size specimens to determine the Master Curve reference temperature, T0, for as‐received and degraded ferritic RPV materials. The usefulness of correlating the Master Curve reference temperature, T0, with the constraint parameter, Q, has been demonstrated; however, the trend curves derived require further development and validation before they can be used in fracture analyses. The output from VOCALIST has contributed in providing the validation of methodology necessary to underpin the diffusion of constraint‐based fracture mechanics arguments in RPV safety cases, with potential applications including WWER as well as Western‐style LWR reactor types.

Analysis of the rejuvenation performance of hybrid blankets by using uranium fuels (UN, UC, UO 2, U 3Si 2) and different coolants for various volume fraction

The possibility of nuclear fuel rejuvenation in fusion reactors is investigated for different fuels and coolants. Neutronic performances of the deuterium–tritium (D–T) driven hybrid blankets, fuelled with UN, UC, UO 2 and U 3Si 2, in four different cases, are investigated under first wall load of the 5 MW/m 2. The fissile fuel zone considered to be cooled with four coolants: air, flibe (Li 2BeF 4), natural lithium and eutectic lithium (Li 17Pb 83) with volume fraction ratio of 29.5, 45.5 and 62.56%. The behaviour of the fuels mentioned above are observed during 48 months for discrete time intervals of Δt =15 days and by a plant factor (PF) of 75%. At the end of the operation time, calculations have shown that cumulative fissile fuel enrichment (CFFE) values have varied between 3.80 and 8.1% depending on the fuel, volume fraction and coolant type. The best enrichment performance is obtained in flibe (Li 2BeF 4) coolant blankets, followed by Eutectic lithium (Li 17Pb 83), air whereas natural lithium coolant shows a poor rejuvenation performance in all fuels. CFFE reach maximum value (8.1%) in UO 2 fuelled blanket (in Row #1) and Li 2BeF 4 coolant that volume fraction is 62.5% after 48 months. The lowest CFFE value (3.80%) is in U 3Si 2 fuelled blanket (in Rows #6 and 7) and natural lithium coolant that volume fraction is 62.56% at the end of the operation period.The enrichment would be sufficient for LWR reactor. The best tritium breeding ratio (TBR) is obtained in U 3Si 2 fuelled blanket with natural lithium coolant, and followed by UC, UO 2, UN with the same coolant. At the beginning of the operation, TBR values were 1.459, 1.502 and 1.554 in U 3Si 2 fuelled blanket with natural lithium coolant 1.414, 1.474 and 1.547 in UC fuelled blanket with natural lithium coolant for volume fraction of 29.5, 45.5 and 62.56%, respectively. At the end of the operation, TBR reach 1.511, 1.559 and 1.613 in U 3Si 2 fuelled blanket and 1.467, 1.532 and 1.609 in UC fuelled blanket for volume fraction of 29.5, 45.5 and 62.56%, respectively. TBR values are higher than unity. Therefore, investigated hybrid blanket is self-sufficient for all fuel mixture and coolants. The isotopic percentage of 240Pu is higher than 5% in all modes with flibe coolant, so that the plutonium component in these modes can never reach a nuclear weapon grade quality during the operation period. This is a very important safety factor. The isotopic percentage of 240Pu is lower than 5% in all blanket with air, natural lithium, and eutectic lithium coolant. In these modes, operation period must be increased for safety.

Potential of a fusion–fission hybrid reactor using uranium for various coolants to breed fissile fuel for LWRs

In this study, neutronic performances of the (D,T) driven hybrid blankets, fuelled with UC2 and UF4, are investigated under first wall load of 5 MW/m2. The fissile fuel zone is considered to be cooled with three coolants: gas (He or CO2), flibe (Li2BeF4), and natural lithium. The behaviour of the UC2 and UF4 fuels are observed during 48 months for discrete time intervals of Δt=15 days and by a plant factor of 75%. At the end of the operation time, calculations have shown that Cumulative Fissile Fuel Enrichment (CFFE) values varied between 5 and 8.5% depending on the fuel and coolant type. The best enrichment performance is obtained in UF4 fuelled blanket with flibe coolant, followed by gas and natural lithium coolant. CFFE reaches maximum value (8.51%) in UF4 fuelled blanket (in row #1) and flibe coolant mode after 48 months. The lowest CFFE value (4.71%) is in UC2 fuelled blanket (in row #8) and natural lithium coolant at the end of the operation period. This enrichment would be sufficient for LWR reactor. At the beginning of the operation, tritium breeding ratio (TBR) values were 1.090, 1.3301 and 1.2489 in UC2 fuelled blanket and 1.0772, 1.2433 and 1.1533 in UF4 fuelled blanket for flibe, natural lithium and gas coolant, respectively. At the end of the operation, TBR reach 1.1820, 1.3983 and 1.3138 in UC2 fuelled blanket and 1.2041,1.3266 and 1.2407 in UF4 fuelled blanket for flibe, natural lithium and gas coolant, respectively. Nuclear quality of the plutonium increases linearly during the operation period. The isotopic percentage of 240Pu is higher than 5% in UF4 and UC2 fuel with flibe coolant, so that the plutonium component in these modes can never reach a nuclear weapon grade quality during the operation period. This is very important factor for safeguarding. The isotopic percentage of 240Pu is lower than 5% in UC2 fuel with gas and natural lithium coolant. In these modes, operation period must be increased to safeguarding.

Neutronic evaluation of the non-proliferating reprocessed nuclear fuels in pressurized water reactors

Alternative reprocessing methods with low grade decontamination and uranium—plutonium coextraction characteristics are proposed for increasing proliferation resistance and permitting recovered fissile materials utilization in LWR reactors. This work brings out a preliminary neutronic study of fuels reprocessed by Coprocessing and AIROX techniques.

Numerical simulation of the inlet plenum flow field for a LWR reactor pressure vessel

FLOW3D has been employed to simulate the turbulent flow occurring in the reactor pressure vessel (RPV) inlet plenum of the simplified boiling water reactor (SBWR) nuclear plant. Calculations have been carried out for a 90° sector of the RPV lower plenum, from the bottom to the core plate level, including downcomer and control rod guide tubes (CRGTs) simulation. A structured body-fitted grid is used to model geometrical details inside the plenum (CRGTs bundle and downcomer). Grid sizes of 162,000 nodes with the k-ε turbulence model have been used. In the present analysis, the usual porosity approach has not been used, while cross flow conditions through CRGTs bundle are taken into account by direct simulation of the tubes gap. Predictions of the steady-state pressure and velocity field have been made. The effects of the number of grid points and boundary conditions on the problem solution have been investigated. Results show that a CFD analysis is able to add new relevant information on the flow pattern inside the plenum, which is usually not taken into account by traditional engineering practice.

Description of the German Concept of Final Storage Cask

AbstractThe German storage concept for the direct final storage of spent fuel assemblies from LWR reactors is described. The final storage concept is designed in such a way that it encompasses the whole spectrum of fuel elements to be stored from German reactors, Le. U fuel assemblies and MOX fuel assemblies with a mean bumup of 55 GW.d.t−1 heavy metal were considered. The further design requirements are defined in such a way that the cask concept satisfies the conditions for type B(U) transport, interim storage and fmal storage. The safe long-term containment of the activity is guaranteed by an inner cask welded leak-tight; the sufficient shielding and the transport packaging are ensured by a shielding cask.

Review of fracture toughness requirements and data relevant to LWR reactor pressure vessels

A survey of current code requirements (ASME and RCC-M) relating to the fracture toughness properties of the ferritic materials used in the fabrication of PWR reactor pressure vessels is presented, together with supporting test data published in the period 1970 to 1985. These fracture toughness properties are discussed in relation to the critical sizes of defect applicable to particular regions of a PWR primary vessel for given design basis transient conditions.

Hydrodynamics of degraded core cooling

The literature on hydrodynamics of forced convection cooling of particle beds is reviewed and used to assess the characteristics of in-situ cooling of a degraded LWR reactor core under conditions representative of severe accidents. It is found that the pressure head required, for a given rate of liquid water flow, through a totally degraded core is one to two orders of magnitude higher than the case of intact core geometry. To remove decay heat of up to 1–2% of the reactor normal power, even with conservative assumptions, the pressure head is within the capability of the main reactor pumps. However, particles with very small diameter (less than 100 μm) will potentially be swept out by the flow. The sensitivity of the hydraulic characteristics to the allowed coolant temperature rise across the fragmented core is also investigated.

Applications of probabilistic fracture mechanics to light water reactor pressure vessels and piping

This paper reviews recent calculations of the statistical reliability of LWR reactor vessels and piping. The broad theoretical principles of these calculations are well established and it is therefore possible to compare the physical assumptions made in different calculations. Such a comparison shows that certain functions are not known at all well; for example, (i) the frequency of occurrence of cracks in weld-regions, (ii) the size distribution of cracks, (iii) the efficiency of methods of non-destructive examination and (iv) the transient loadings that the system experiences in service. On the other hand, relevant materials properties (toughness, crack growth characteristics) appear to be known adequately if not completely. Despite these quantitative uncertainties in the input, it seems possible to draw several broad conclusions from the results of these calculations. These concern (i) the low absolute rates of failure, (ii) the way these depend upon time in service, (iii) the effect upon them of in-service inspection and (iv) their sensitivity or otherwise to the physical assumptions which are made.

The statistical fuel model, FRP

The reliability of nuclear fuel for LWR reactors is generally good, but there are still some failures caused by pellet-cladding interactions (stress corrosion). Only a very small fraction of the fuel pins in operation fail, and the nature of the failures indicates a statistical problem; therefore a statistical fuel model was developed. The statistical methods are either Monte Carlo simulation or a Taylor approximation. The statistical program utilizes a deterministic fuel performance code with a stress corrosion failure criterion, verified against experimental data. The failure probability for a power ramp was evaluated for different fuel designs and ramp conditions. It is shown that the failure probability can be quite large, even when a deterministic simulation (based on the average values for design and material data) is far from the failure criterion.

Advanced HTGR systems

G A = Y = NOMENCLATURE ratio of thermal reactor to breeder thermal power ratio of thermal breeder to LWR reactor thermal p o w e r conversion ratio total fissions/thermal fissions in any reactor out-of-pile inventory/in-pile inventory capture/fission ratio fissile rating [MW(t)] per kg fissile fissile out]fissile in ratio in loadings fissions per initial fissile atom burnup MWD per tonne metal enrichment annual fractional industry growth rate fuel life years L = load factor Suffixes B = breeder L = LWR None = HTGR 1. B A C K G R O U N D The prime objective of the H T G R system was, and still is, to provide nuclear heat at a high temperature. This is an objective of great merit, not just for the obvious efficiency advantage to any associated heat engine but because it also offers prospect of applications not otherwise open to use of nuclear heat at all, including association with closed cycle gas turbines and supplying high temperature process heat, for which fuel scarcity and environmental cleanliness considerations are creating increasing demand. While such attractions were in mind from the start, the prudence of dealing with one set of development problems at a time and the desirability of a more immediately needed task were also appreciated. Firm commercial establishment of the H T G R for steam-powered electric generation was in fact the obvious first goal. The most immediately realizable reward for high temperature in this field was the ability to supply the steam conditions already adopted for contemporary fossil plants, and thus to partake of the cost benefits accruing from their extensively evolved and standardized machinery. The required 1000°F steam temperature, however, called for helium temperatures of no more than 1300°-1400°F, the further reduction in boiler surface offered by higher temperatures being more than offset by necessary specialty material cost. The essential point here is that this step far from fully exploited the inherent capabilities of the all-graphite type of core which principally characterizes the HTGR. It did, however, incur the main cost penalty that this form of core construction inevitably involved, which was that it required a fabricated moderator which had to be disposed of with every loading. This consideration accentuated the need for extracting the highest possible energy per unit mass of fuel-graphite elements. The moderator-diluted fuel and graphite structure of H T G R fuel elements already conferred to them ability to survive long burnup (~105 MWD/tonne), but to explo!t this fully it would be necessary to match this with a correspondingly long period of sustained reactivity. This is one reason why the thorium cycle was introduced from the start, because the gain in conversion ratio associated with the high neutron yield available from the 2zzU generated by this cycle greatly aids reactivity maintenance. There was also, of course, a more general motivation to tap the additional fuel resources represented by thorium, proper exploitation of which calls moreover for the high conversion ratios which were available. An important point here is that, in the absence of coexisting high gain fast breeders (the only practically realizable external source of °-~U-), the HTGP, of necessity has to use zasU for its initial fissile inventory. This handicaps initial nuclear performance and introduces dependence on a near fully enriched uranium supply, raising both the cost of separation and the amount of natural fissile supply necessarily discarded in separation plant tails. This places a premium on either attaining a good conversion ratio even in the initial H T G R core or seeking the independent source of z~U that could come from fast reactor breeding blankets. Even starting with z35U, the H T G R could be designed for a conversion ratio approaching 0.85. Such a core would accumulate zmU rapidly and fully enough to very effectively exploit the thorium2z3U cycle after a few refuelings. By contrast, a conversion rato of only 0.6 or so would still entail a perpetual dependence on a highly separated ~asU supply, for at this level the buildup of the zazU content, even with infinitely repeated recycling, would amount to only a fraction of the total fission content.