Lead Coolant Test Facility—Design concept and requirements

The Idaho National Laboratory prepared a preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a Lead Coolant Test Facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic. Based on review of current world lead or lead-bismuth test facilities and research need listed in the Generation IV Roadmap, five broad areas of requirements are identified in this paper:

•

Develop and demonstrate feasibility of submerged heat exchanger

•

Develop and demonstrate open-lattice flow in electrically heated core

•

Develop and demonstrate chemistry control

•

Demonstrate safe operation

•

Provision for future testing

Across these five broad areas are supported by twenty-one specific requirements. The purpose of this facility is to focus the lead fast reactor community domestically on the requirements for the next unique state of the art test facility. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 420 °C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7 M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.     

Thermal-hydraulics analyses of ELSY lead fast reactor with open square core option

Among the next generation of nuclear reactors, well-known as Gen-IV, the LFR (Lead Fast Reactor) is one of the most promising advanced reactor able to comply the principles of sustainability, economics, safety and proliferation resistance. The ELSY project (European Lead-cooled SYstem), funded by the 6th European Framework Programme, aims at investigating the technical/economical feasibility of a high power lead fast reactor with waste transmutation capability.Several innovative design solutions have been proposed at system level and some of them regard the core region with open-assemblies that represents the reference option. To support the design phase and the safety assessment of ELSY, the thermal-hydraulic code RELAP5, modified to treat heavy liquid metals, has been taken into account.The paper deals with the development of the ELSY thermal-hydraulic and point kinetic model for RELAP5 focusing the attention at core assembly level to verify that the temperatures at nominal power conditions stay within the safety limits both in Beginning-of-Cycle (BOC) and in End-of-Cycle (EOC) conditions. Moreover, in order to have a first evaluation of the system behavior in accidental conditions, an Unprotected Loss-of-Flow Accident (ULOF) simulation at BOC has been analysed and discussed.

List of acronyms

ADS

Accelerator Driven System

BOC

Beginning-of-Cycle

DEC

Design Extension Condition

EFIT

European Facility for Industrial Transmutation

ELSY

European Lead-cooled SYstem

EOC

End-of-Cycle

FA

Fuel Assembly

FP6

6th Framework Programme

Gen-IV

Generation IV (four)

GESA

Gepulste ElektronenStrahlanlage (Pulsed Electron Beam Facility)

HX

Heat eXchanger

IC

Isolation Condenser

LFR

Lead Fast Reactor

LMFR

Liquid Metal Fast Reactor

PDS-XADS

Preliminary Design Studies on eXperimental ADS

RVACS

Reactor Vessel Air Cooling System

ULOF

Unprotected Loss-of-Flow

W-DHR

Water Decay Heat Removal

     

Anomalies and other concerns related to the critical heat flux

This paper summarizes various unusual trends in the critical heat flux (CHF) that have been observed experimentally in tubes or bundle subassemblies. They include the following:

•

Occurrence of a minimum in the CHF vs. quality (X) curve at high flows - leading to an initial upstream CHF occurrence in uniformly heated channels. This phenomenon has been observed at high flows in both water and Freon.

•

Occurrence of a limiting quality region on the CHF vs. X curve where the CHF drops by 30-90% for a nearly constant quality. This is thought to correspond to the boundary between the entrainment controlled and the deposition controlled region and causes problems for prediction methods of the form CHF = f(X).

•

Impact of flow obstructions on the occurrence of upstream CHF and the limiting quality region. The additional mixing by grid spacers or bundle appendages results in a more homogeneous phase distribution, and diminishes the effects of flow regime/heat transfer regime transitions responsible for some of the unusual CHF trends, and results in a more gradually decreasing CHF vs. X curve.

•

Absence of a CHF temperature excursion at high flows and high qualities - this is found to be caused by a change in slope of the transition boiling part of the boiling curve from a negative value (usual trend that results in a temperature excursion) to a positive slope.

•

Gradual disappearance of the sharp temperature excursion at CHF when increasing the pressure towards and beyond the critical pressure - no drastic change is observed in the axial temperature distribution of a heated tube experiencing CHF when, for constant mass flux and inlet temperature, the pressure is gradually increased from subcritical to supercritical.

•

CHF fluid-to-fluid modelling: differences in CHF trends at certain conditions between refrigerants and water at equivalent conditions.

The mechanisms responsible for these trends and the implications for bundle geometries are discussed.Concerns regarding the reported uncertainty of predicted CHF values and the range of application of CHF prediction methods are also discussed.     

Plastic collapse and LBB behavior of statically indeterminate piping system subjected to a static load

The plastic collapse and LBB behavior of statically indeterminate piping system were investigated in this study, compared with those of the statically determinate piping system. Special attention was paid to evaluate the crack opening displacement after a crack penetrated wall thickness. The main results obtained were as follows:

1. The reduction of ultimate strength caused by a crack was relatively small in the statically indeterminate piping system. The main reason is thought to be that a sufficient redistribution of the bending moment occurs in this system.

2. A method to evaluate the crack opening displacement after crack penetration in a pipe with a non-penetrating crack was proposed. From this method, it was known that the crack opening displacement could be evaluated by using the incremental plastic rotation angle.

3. The acceptable defect size considering the deformation of a pipe was estimated by comparing the plastic moment at the defective part and the gross yielding moment at the non-defective part.

Cosmic ion bombardment of the icy moons of Jupiter

A large number of experiments have been performed in many laboratories in the world with the aim to investigate the physico-chemical effects induced by fast ions irradiating astrophysical relevant materials. The laboratory in Catania (Italy) has given a contribution to some experimental works. In this paper I review the results of two class of experiments performed by the Catania group, namely implantation of reactive (H⁺, C⁺, N⁺, O⁺ and S⁺) ions in ices and the ion irradiation induced synthesis of molecules at the interface between water ice and carbonaceous or sulfurous solid materials. The results, discussed in the light of some questions concerning the surfaces of the Galilean moons, contribute to understand whether minor molecular species (CO₂, SO₂, H₂SO₄, etc.) observed on those objects are endogenic i.e. native from the satellite or are produced by exogenic processes, such as ion implantation.The results indicate that:

-

C-ion implantation is not the dominant formation mechanism of CO₂ on Europa, Ganimede and Callisto.

-

Implantation of sulfur ions into water ice produces hydrated sulfuric acid with high efficiency such to give a very important contribution to the sulfur cycle on the surface of Europa and other satellites.

-

Implantation of protons into carbon dioxide produces some species containing the projectile (H₂CO₃, and O-H in poly-water).

-

Implantation of protons into sulfur dioxide produces SO₃, polymers, and O₃ but not H-S bonds.

-

Water ice has been deposited on refractory carbonaceous materials: a general finding is the formation of a noteworthy quantity of CO₂. We suggest that this is the primary mechanism to explain the presence of carbon dioxide on the surfaces of the Galilean satellites.

-

Water ice has been deposited on refractory sulfurous materials originating from SO₂ or H₂S irradiation. No evidence for an efficient synthesis of SO₂ has been found.

     

SEMER: a simple code for the economic evaluation of nuclear and fossil energy-based power production systems

The code system, SEMER, was recently developed to evaluate the economic impact of various nuclear reactors and associated innovations. Models for nearly all fossil energy-based systems were also included to provide a basis for cost comparisons.Essentially, SEMER includes three types of model libraries: the global model, for a rapid estimation of various nuclear and fossil energy-based systems, the detailed models, for the finer cost evaluation of individual components and circuits in a PWR type of reactor and the fuel cycle models, for PWRS, HTRs and FBRs, allowing the cost estimations related to all the steps in the nuclear fuel cycle, including reprocessing and disposal.This paper summarises our on-going investigations on new developments in, and on the validation of, the SEMER system.Details of the modelling principles, and the results of validation carried out in the context of an EDF/CEA Joint Protocol Agreement, are also presented.First results of this validation are highly encouraging:

• Relative errors for the total kWh or overnight and investment costs are less than 5% for large PWR systems operating in France or other countries.

• These errors are less than 3% for small-sized compact PWRs and they are of the order of 4–7% for HTRs (as compared to IAEA estimations).

• For fossil energy-based power plants, the relative error, even with slightly different cost breakdown between SEMER and that of existing installations, is from 5 to 20%.

• Similarly, errors on the nuclear fuel cycle costs are about 1–4%, compared to published reference values.

On the occurrence of burnout downstream of a flow obstacle in boiling two-phase upward flow within a vertical annular channel

If a flow obstacle, such as a spacer is placed in a boiling two-phase flow within a channel, the temperature on the surface of the heating tube is severely affected by the existence of the spacer. Under certain conditions, a spacer has a cooling effect, and under other conditions, the spacer causes dryout of the cooling water film on the heating surface. The burnout mechanism, which always occurs upstream of a spacer, however, remains unclear.In a previous paper [Fukano, T., Mori, S., Akamatsu, S., Baba, A., 2002. Relation between temperature fluctuation of a heating surface and generation of drypatch caused by a cylindrical spacer in a vertical boiling two-phase upward flow in a narrow annular channel. Nucl. Eng. Des. 217, 81–90], we reported that the disturbance wave has a significant effect on dryout and burnout occurrence and that a spacer greatly affects the behavior of the liquid film downstream of the spacer.In the present study, we examined in detail the influences of a spacer on the heat transfer and film thickness characteristics downstream of the spacer by considering the result in steam–water and air–water systems. The main results are summarized as follows:

(1) The spacer averages the liquid film in the disturbance wave flow. As a result, dryout tends not to occur downstream of the spacer. This means that large temperature increases do not occur there. However, traces of disturbance waves remain, even if the disturbance waves are averaged by the spacer.

(2) There is a high probability that the location at which burnout occurs is upstream of the downstream spacer, irrespective of the spacer spacing.

(3) The newly proposed burnout occurrence model can explain the phenomena that burnout does occur upstream of the downstream spacer, even if the liquid film thickness t_F m is approximately the same before and behind the spacer.