Shutdown and Closure of the Experimental Breeder Reactor–II

The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and maintenance requirements during the interim period between deactivation and decommissioning. The plans also establish document archival of not only all the closure documents, but also the key plant documents (P&IDs, design bases, characterization data, etc.) in a convenient location to assure the appropriate knowledge base is available for decommissioning, which could occur decades in the future.

A Study on the Embedment Effect in the Soil-Structure Interaction Analysis of the APR1400

The SSI (Soil-Structure Interaction) analyses are being performed for the APR1400 (Advanced Power Reactor 1400MWe, Old name - KNGR ; Korean Next Generation Reactor) design, because the APR1400 is developed as a Standard Nuclear Power Plant concept enveloping suitable soil conditions. For the SSI analyses, SASSI program which adopts the Flexible Volume Method is used. In the SSI analyses, there can be uncertainties by Bond and De-bond problem between the structure and lateral soil elements. According to ASCE Standard 4, one method to address this concern is to assume no connectivity between structure and lateral soil over the upper half of the embedment of 20ft (6m), whichever is less. This study is performed as a part of the parametric analyses for the APR1400 seismic analyses to address the concern of the potential embedment effect on the in-structure response spectra due to connectivity between structure and lateral soil. In this study, 4 model cases are analyzed to check the potential embedment effect — Full connection, 20ft no connectivity which is defined as a minimum De-bond depth of the soil in ASCE Standard 4 and 26.5ft no connectivity between structure and lateral soil over the upper half of the embedment. Last one is full no connection for only reference. The in-structure response spectra are compared with the response spectra without considering the embedment effect.

Status of the United States Nuclear Regulatory Commission Pressurized Thermal Shock Rule Re-Evaluation Project

The current federal regulations to ensure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models that were developed in the early to mid 1980s. Since that time, there have been advancements in relevant technologies associated with the physics of PTS events that impact RPV integrity assessment. Preliminary studies performed in 1999 suggested that application of the improved technology could reduce the conservatism in the current regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license extension considerations. Based on the above, in 1999, the United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project, with the nuclear power industry as a participant, to re-evaluate the current PTS regulations within the framework established by modern probabilistic risk assessment (PRA) techniques. During the last three years, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, PRA, human reliability analysis (HRA), materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. These experts were from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have now been implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions. The baseline version of FAVOR (version 1.0) was released in October 2001. The updated risk-informed computational methodology in the FAVOR code is currently being applied to selected domestic commercial pressurized water reactors to evaluate the adequacy of the current regulations and to determine whether a technical basis can be established to support a relaxation of the current regulations. This paper provides a status report on the application of the updated computational methodology to a commercial pressurized water reactor (PWR) and discusses the results and interpretation of those results. It is anticipated that this re-evaluation effort will be completed in 2002.

Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of the AMPs, which do not require further evaluation, do need enhancements to allow for an extended period of reactor operation. The paper concludes that the GALL report systematically evaluates the generic AMPs for their effectiveness during an extended period of operation and provides a technical basis for their acceptance. The use of the GALL report should facilitate both preparation of a license renewal application and timely and uniform review by the NRC staff.

Load Regime Diagram of a Pipe With Cyclically Plastic Bending

The criteria of piping seismic design based on linear elastic analysis has been proved to be conservative, which is mainly because the influence of plastic deformation on piping dynamic response is neglected. In the present paper, a pipe under seismic excitation is simplified as an beam with tubular cross section subjected to steady axial force and fully reversed cyclic bending moment, and the elastic-plastic behavior of the pipe is studied. Various behavior of the pipe under different combinations of axial force and cyclic bending moment is discussed and the boundary curve equations between them are obtained. Also the load regime diagram for a pipe which is formed by the boundary curve equations in the loading plane is given, from which the elastic-plastic behavior of the pipe can be determined directly.

Structural Integrity Assessment of Steam Generator Tubes Using a New EPRI Statistical Approach

A fundamental step in tube plugging management of a Steam Generator (SG), in a Nuclear Power Plant (NPP), is the tube structural integrity evaluation. The degradation of SG tubes may be considered one of the most serious problems found in PWRs operation, mainly when the tube material is the Inconel 600. The first repair criterion was based on the degradation mode where a uniform tube wall thickness corrosion thinning occurred. Thus, a requirement of a maximum depth of 40% of the tube wall thickness was imposed for any type of tube damage. A new approach considers different defects arising from different degradation modes, which comes from the in-service inspections (NDE) and how to consider the involved uncertainties. It is based on experimental results, using statistics to consider the involved uncertainties, to assess structural limits of PWR SG tubes. In any case, the obtained results, critical defect dimensions, are within the regulatory limits. In this paper this new approach will be discussed and it will be applied to two cases (two defects) using typical data of SG tubes of one Westinghouse NPP. The obtained results are compared with ‘historical’ approaches and some comments are addressed from the results and their comparison.

Effects of Dynamic Strain-Aging and Cyclic Loading on Fracture Behavior of Ferritic Steels

Ferritic steels commonly used for pressure vessels and reactor supports in light water reactors (LWRs) exhibit dynamic strain aging (DSA) resulting in decreased ductility and toughness. In addition, recent work indicated decreased toughness during reverse-cyclic loading that has implications on reliability of these structures under seismic loading conditions. We summarize some of our recent work on these aspects along with synergistic effects, of interstitial impurity atoms (IIAs) and radiation induced point defects, that result in interesting beneficial effects of radiation exposure at appropriate temperature and strain-rate conditions. Radiation-defect interactions were investigated on pure iron, Si-killed mild steel, A533B, A516, A588 and other reactor support and vessel steels. In all cases, DSA is seen to result in decreased ductility accompanied by increased work-hardening parameter. In addition to mechanical property tests, fracture toughness is investigated on both A533B and A516 steels. While dips in fracture toughness are observed in A533B steel in the DSA region, A516 steel exhibited at best a plateau. The reasons could lie in the applied strain-rates; while J1c tests were performed on A533B steel using 3-point bend tests on Charpy type specimens, CT specimens were used for A516 steel. However, tensile and 3-point bend tests on similar grade A516 steel of different vintage did exhibit distinct drop in the energy to fracture. Load-displacement curves during J1c tests on CT specimens did show load drops in the DSA regime. The effect of load ratio (R) on J versus load-line displacement curves for A516 steel is investigated from +1 to −1 at a fixed normalized incremental plastic displacement of 0.1 (R = 1 corresponds to monotonic loading). We note that J-values are significantly reduced with decreasing load ratio. The work-hardening characteristics on the fracture surfaces were studied following monotonic and cyclic loading fracture tests along with the stress-field analyses. From the hardness and the ball-indentation tests, it was shown that decreased load ratio (R) leads to more strain hardening at the crack tip resulting in decreased fracture toughness. From the stress field analysis near the crack tip of a compact tension fracture toughness test specimen, a cycle of tensile and compressive loads is seen to result in tensile residual stresses (which did not exist at the crack tip before). These results are important to evaluations of flawed-structures under seismic loading conditions, i.e. Leak-Before-Break (LBB) and in-service flaw evaluation criteria where seismic loading is addressed. In addition, studies on fast vs total (thermal+fast) neutron spectra revealed unexpected results due to the influence of radiation exposure on source hardening component of the yield stress; grain-size of pure iron plays a significant role in these effects.

Increased Availability From Improved Condenser Design

Performance parameters and flow characteristics on the shell side of surface condensers are becoming better understood. Contributing to this knowledge base is the recent ability to measure the physical properties as well as the quantity of gases being removed from the condenser by air removal equipment. Reviewed here are the commonality of these data from many operating condensers obtained over the past six years and other known condenser measurements, theory and laboratory experiments. These are combined to formulate global theoretical description of condenser dynamics describing the mechanism responsible for aeration and de-aeration, excess back pressure buildup due to air ingress or generation of other noncondensable gases, and the dissolubility of corrosive gases in condensate. The theoretical description supports a dynamic model useful for deciding condenser configuration design and design improvements. Features of design found in many operating condensers that promote aeration and resulting corrosion are presented. The benefits of the model and engineering design modifications to plant life cycle management, improved condenser performance, outage reduction and reliability improvements, lost load recovery and fuel savings are discussed.

A Contribution to Proof the Component Integrity Taking Into Account the Corrosion-Assisted Crack Growth

The proof of the component integrity is fundamental for a safe and reliable operation of Nuclear Power Plants (NPP). The concept of the Material Testing Institute (MPA) for integrity assessment is based on fracture mechanic analysis which results in detailed regulations for nondestructive examination. This approach has to account for the main damage mechanisms as fatigue and corrosion. This paper focuses on the influence of corrosion-assisted crack growth which strongly depends on corrosion and environmental conditions (e.g. coolant purity). Up to stress intensity of approximately 60 MPa√m for ferritic low-alloy steels in high-purity water (acc. to specification) under constant load conditions the analysis can be based on a crack extension of max. 70 µm for each load cycle. Related to a test duration of 1000 hours this is equivalent to a formally calculated crack growth rate (CGR) of = 2 · 10−8 mm/s. For austenitic stainless steels more complex dependences on material, environmental and mechanical parameters exist. Particularly, for stabilized austenitic steels the crack growth rate data base is relatively weak. Under unfavourable environmental conditions in single cases crack growth rates up to 6 mm/a have been measured. Based on experimental results an arithmetic mean value of 0.95 mm/a and a median value of 0.6 mm/a have been determined. A further improvement of data base is desirable.

U.S. Operating Experience With Thermally Treated Alloy 600 Steam Generator Tubes

Steam generators placed in service in the 1960s and 1970s were primarily fabricated from mill-annealed Alloy 600. Over time, this material proved to be susceptible to stress corrosion cracking in the highly pure primary and secondary water chemistry environments of pressurized-water reactors. The corrosion ultimately led to the replacement of steam generators at numerous facilities, the first U.S. replacement occurring in 1980. Many of the steam generators placed into service in the 1980s used tubes fabricated from thermally treated Alloy 600. This tube material was thought to be less susceptible to corrosion. Because of the safety significance of steam generator tube integrity, this paper evaluates the operating experience of thermally treated Alloy 600 by looking at the extent to which it is used and recent results from steam generator tube examinations.