Study on Chemical Reactivity Control of Liquid Sodium: Development of Nano-Fluid and Its Property and Applicability to FBR Plant

Liquid sodium is used as the coolant of the fast breeder reactor (FBR). A weak point of sodium is a high chemical reactivity with water or oxygen. So an idea of chemical reactivity suppression of liquid sodium itself is proposed. The idea is that nano-meter size particles (hereafter called nano-particles) are dispersed in liquid sodium, and an atomic interaction which is generated between nano-particle and sodium atoms is applied to suppress the chemical reactivity. We call sodium that has dispersed the nano-particle a Nano-fluid. Three key technologies which are the trial manufacture of Nano-fluid, the reaction property of the Nano-fluid and applicability of Nano-fluid to FBR Plant have been carried out to develop the Nano-fluid.

Development of Low-Activation Design Method for Reduction of Radioactive Waste Below Clearance Level

Design methodology for reinforced concrete of nuclear power plants to reduce radioactive wastes in decommission phase has been developed. To realize this purpose, (1) development of raw materials database of cements, aggregates and steel bars on concentration of radioactive target elements, (2) trial production of low activation cements and steel bars based on the material database developed in (1), and (3) development of tools for estimation and prediction of the amount of radioactive elements in reactor shielding walls have been carried out. Radioactive analysis showed that Co and Eu were the major target elements which decide the radioactivity level of reinforced concrete from wide survey of raw materials for concrete (typically aggregates and cements). Material database for the contents of Co and Eu was developed based on the chemical analysis and radioactivation analysis. Upon the above survey and execution expreiment of concrete, six types of low-activation concrete are proposed for various radioactive portion in the plant. These concrete have a 1/10 – 1/300 rasioactivity compare to the ordinary concrete, which are assumed the concrete with Andesite aggregate and ordinary Portland cement. Baed on the above data base, it was clarified that the low activation cement would be successfully manufactured by adequate selection of raw materials. The prospect to produce the low-heat portland cement which would have a 1/3 radioactivity in comparison with conventioanl cements obtained by means of selection of limestone and natural gypsum. An attempte was carried out to produece low activation heavy-mortar which would have radioactivity below the clearance level when using at the radiation shielding wall of BWR. Characterization and optimization of consturction conditions with new additives have also been carried out. These two new raw materials for low-activation concrete are conducted in pre-manufacture size, and over the laboratry level. Boron added low-activation concrete are also carried out as extreamly high performance low-activation concrete. It was claryfied that the accurcy of calculation results of the radioactivity evaluation was very high compared to available benchmark calculation for the JPDR and commercial light water reactor. The specification of the mapping system for judging the activation classification was also developed by using the general-purpose radio activation calculation tool. This work is supported by a grant-in-aid of Innovative and Viable Nuclear Technology (IVNET) development project of Ministry of Economy, Trade and Industry, Japan.

Development of Code Case N-755: Use of Polyethylene (PE) Plastic Pipe for Section III, Division 1, Construction and Section XI Repair/Replacement Activities

With the increasing failures of metallic pipe in nuclear Service Water Systems, a new material needed to be found. One option is polyethylene (PE) pipe. PE pipe can be used in non-safety applications at a nuclear plant using the American Society of Mechanical Engineers (ASME) B31, Standards of Pressure Piping with no regulatory review. However, the use of PE material in safety applications, which are regulated by the Nuclear Regulatory Commission (NRC), necessitates a new Standard with special requirements. At the request of the Duke Power Corporation, a new ASME Standard was written by a special Project Team. This standard is found in the form of a Code Case under the control of the ASME Boiler & Pressure Vessel Code (B&PVC). The Code Case utilizes Sections of the B&PVC as its foundation and includes the design, procurement, installation, fusing, examination and testing requirements for the use of PE pipe within safety systems. The first version of the Code Case contained only the minimum requirements needed to support Duke Power Corporation’s first phase of PE piping installation into a safety system within a nuclear power plant. The Code Case developed is titled, N-755, Use of Polyethylene (PE) Plastic Pipe for Section III, Division 1, Construction and Section XI Repair/Replacement Activities. The first version of this case is limited to buried piping using only the following components; straight PE pipe, PE mitered elbows, and transition flanges. The Code Case will be revised as data for material and components becomes available at the completion of testing.

Non-Linear Design Verification of Nuclear Power Piping According to ASME III NB/NC

A nuclear piping system which is found to be disqualified, i.e. overstressed, in design evaluation in accordance with ASME III, can still be qualified if further non-linear design requirements can be satisfied in refined non-linear analyses in which material plasticity and other non-linear conditions are taken into account. This paper attempts first to categorize the design verification according to ASME III into the linear design and non-linear design verifications. Thereafter, the corresponding design requirements, in particular, those non-linear design requirements, are reviewed and examined in detail. The emphasis is placed on our view on several formulations and design requirements in ASME III when applied to nuclear power piping systems that are currently under intensive study in Sweden.

Recirculation Pump Speed Control System Upgrades in Nuclear Boiling Water Reactors

The nuclear power industry is faced with determining what to do with equipment and instrumentation reaching obsolescence and selecting the appropriate approach for upgrading the affected equipment. One of the systems in a nuclear power plant that has been a source of poor reliability in terms of replacement parts and control performance is the reactor recirculation pump speed/ flow control system for boiling water reactors (BWR). All of the operating BWR-3 and BWR-4’s use motor-generator sets, with a fluid coupled speed changer, to control the speed of the recirculation water pumps over the entire speed range of the pumps. These systems historically have had high maintenance costs, relative low efficiency, and relatively inaccurate speed control creating unwanted unit de-rates. BWR-5 and BWR-6 recirculation flow control schemes, which use flow control valves in conjunction with two-speed pumps, are also subject to upgrades for improved performance and reliability. These systems can be improved by installing solid-state adjustable speed drives (ASD), also known as variable frequency drives (VFD), in place of the motor-generator sets and the flow control valves. Several system configurations and ASD designs have been considered for optimal reliability and return on investment. This paper will discuss a highly reliable system and ASD design that is being developed for nuclear power plant reactor recirculation water pump controls. Design considerations discussed include ASD topology, controls architecture, accident, transient and hydraulic analyses, potential reactor internals modifications, installation, demolition and economic benefits.

Development of Atlas of Defects of Steam Generators Heat Exchanging Tubes of NPPs With WWER and Analysis of Testing Results

The basic method of the operational inspection of metal of heat exchanging tubes (HET) of steam generators (SG) is eddy-current multifrequency method all over the world. The greatest distribution was received variant with eddy-current testing (ECT) by use of a bobbing probe on the Russian nuclear power plants (NPP). Tubes with the defects which a subject to plugging are defined by results of lead operational ECT over the certain algorithm. SG resource is settled and replacement is required at achievement of a limit on number of the plugged tubes [1].

Developing a Standard Code for Spent Fuel and High Level Waste Containment Baskets

ASME Section III, Division 3, “Containments for Transportation and Storage of Spent Nuclear Fuel and High Level Radioactive Material and Waste” currently addresses the design of transportation and storage containment shells but it has yet to address the containment internal support structure that holds the spent fuel or high level waste in place. However, the code for internal support structures, hereafter referred to by its common name “basket”, has been under development by ASME for the past 2 years. Development of the new code, to be known as Subsection WD, “Internal Support Structures” was deemed necessary because current containment system basket construction is a piecemeal approach using ASME Section III, Division 1, Subsection NF, “Supports” and/or ASME Section III, Division 1, Subsection NG, “Core Support Structures” or some other engineering method. Approvals for the various combinations are granted from the regulatory authority. The piecemeal approach tries to capture the critical elements important for a containment basket. However, Subsections NF and NG are based on nuclear power plant design which has different design goals than for a spent fuel or high level waste containment. The issuance of Subsection WD will ensure standardization of future containment baskets, assist the regulatory agency in the review and approval of the baskets, and ensure that the essential criteria in the basket related to spent fuel and high level waste storage transportation and disposal is adequately addressed. The purpose of the basket is primarily to ensure that the radioactive components in the containment are supported in a way as not to create a criticality event. Current acceptance is typically based on a no yield design that the containment manufactures all say is too conservative and based on unreasonable criteria. What should the basket design be based on, how should Subsection WD address them, etc.? The purpose of this paper is to inform interested parties of the progress that has been made in development of Subsection WD, what construction provisions it will initially include and what is planned for it, and when is it scheduled to be issued.

Allowable Load Derivation of Single Angle Members for Small Bore Pipe Supports Welded on 4-Bolt CEA Plate in APR1400

Single angle members have been rarely used as supporting structures in nuclear power plant because they are open sections which have significantly reduced capacities when considered in comparison to closed sections, and have weakness for twisting load such as local torsion caused by loading eccentricity of geometric center and shear center. However, in APR1400 (Advanced Power Reactor 1400 MW-class in Korea), the extended application of single angle members for supporting structures of small bore piping systems is considered to enhance the constructability and economics of plant. Furthermore, although it is general guideline for support design in APR1400 that supporting structures for equipment should be directly welded to embedded plates or steel structures in buildings as far as possible, in the case of small bore piping system, for the low level priority of construction in site, supporting steel structures for small bore piping could not be evitable to be welded onto the CEA (Concrete Expansion Anchor) plate. Per the ACI 318, ACI 349 and ACI 355.2, most CEA plate designs and anchor bolt load determinations are now based on finite element models that many applications have been individually made for CEA plates. If single angle members are attached onto these plates, integrated finite element models should be developed and analyzed in detail accroding to NRC IEB 79-02. Such a detailed analysis may appear to be excessive to small bore pipe supports which have diverse design materials and frequently subjected to field changes requiring rapid revision. Consideration should be given to reviewing current practices and reducing the level of effort being used for the integrated analysis of support and CEA plate by developing consensus standard regarding reasonable support and CEA plate designs. In this paper, allowable piping loads for single angle members such as L2×2×1/4, L2-1/2×2-1/2×1/4, L3×3×3/8, and L4×4×1/2 welded on the 4-bolt CEA assembly are derived and reviewed for general use for small bore pipe support design, and L2-1/2×2-1/2×1/4 and L3×3×3/8 welded onto 4-bolt (3/8″Φ sleeve type) CEA plate (1/2″×9″×9″) are recommended as standard small bore pipe supports with post-installed anchor system in APR1400.

Non-Linear BWR Stability Analysis

Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants already during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge oN BWR instabilities. The appropriate representation of the physical processes in the non-linear regime requires typically time domain stability analysis. The objective of this paper is to present a physical model, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. The application of this model gives a deeper insight into the physical reasons for the prevention of the uncontrolled divergence of BWR oscillations. The mechanisms that have a stabilizing effect are demonstrated.

Radiation Protection Technician Two-Year Associates of Applied Science Curriculum for National Implementation

The U.S. Department of Labor awarded a $2.3 million grant to the University of Missouri-Columbia (MU) in 2006 in response to the need for well-trained Radiation Protection Technicians (RPTs). The RPT curriculum initiative resulted from significant collaborations facilitated by MU with community colleges, nuclear power plants, professional organizations, and other nuclear industry stakeholders. The objective of the DOL project is to help increase the pool of well-qualified RPTs to enter the nuclear workforce. Our work is designed to address the nuclear industry’s well-documented, increasingly significant need for RPTs. In response to this need, MU and AmerenUE’s Callaway Nuclear Power Plant first partnered with Linn State Technical College’s Advanced Technology Center (LSTC/ATC) to initiate a two-year RPT degree program. The success of this program (enrollments have been increasing over the past four years to a Fall 2007 enrollment of 23) enabled the successful proposal to the DOL to expand this program nationwide. DOL participants include the following partners: Linn State Technical College with AmerenUE – Callaway; Central Virginia Community College with AREVA; Estrella Mountain Community College with Arizona Public Service – Palo Verde; MiraCosta Community College with Southern California Edison – San Onofre; and Hill College with Texas Utilities – Comanche Peak. The new DOL grant has allowed redevelopment of the LSTC/ATC curriculum using a web-based, scenario driven format, benchmarked against industry training standards. This curriculum will be disseminated to all partners. Integral in this curriculum is a paid, three to four month internship at a nuclear facility. Two of the six new RPT courses have been developed as of the end of 2007. Four of five partner schools are accepting students into this new program starting in the winter 2008 term. We expect that these institutions will graduate 100 new RPTs per year to help alleviate the personnel shortage in this critical area of need.