Asset Management Evaluation: The Key Role of Expert Elicitation

The life management of a nuclear power plant raises several major issues amongst which ranks the aging management of the key components of the plant, both from a technical and an economic point of view. Decision-makers are thus faced with the need to define the best strategy in order to achieve the best possible performance which requires both a very precise modeling of the plant and a detailed analysis of all risks potentially incurred. In this paper, we wish to provide the reader with an overview of how advanced expert elicitation techniques can help identify, structure, quantify and feed sensitive data into a risk-based information system which can then be used for risk-based asset management evaluation. First we focus on the way knowledge management techniques allow EDF to structure and collect life-cycle management data into knowledge-based information systems. The elicitation of component experts is key, particularly in order to get technical information on the future behavior of the component (“anticipation” of whatever events may occur on the plant). We then detail how expert elicitations allow to quantify the probabilities of occurrence of the events: whether there is feedback data, models or not, expert opinion has to be taken into account and mixed, for instance with Bayesian procedures, to this information. Lastly we describe how the information elicited from experts can help top level decision makers get a transverse, long term view on how life management investment strategy translates into plant availability, avoided costs and improved component durability.

Post Fire Transient Temperature Distribution in Drum Type Packages

This study investigates the temperature distribution in an idealized cylindrical package subjected to the HAC Fire transient. Cases for several common overpack materials, with thermal conductivity spanning two orders of magnitude, are considered. The results show that the interior temperature distribution and maximum interior temperature are determined by the heat generation of the contents and the thermal resistance of the package materials. Heat generation has a dominant effect on the peak temperature in the center (containment vessel region) of the package, when the internal thermal resistance is high. For cases where the internal resistance is low, heat conducted into the interior during the fire determines the peak temperature in the center, containment vessel region. The thermal wave effect, where the interior temperature continues to rise after the end of the fire exposure, is present in all cases. The study complements the parametric studies of effects of thermal properties on thermal response of packages which were previously reported.

Drop Simulation of 6M Drum With Locking-Ring Closure and Liquid Contents

This paper presents the dynamic simulation of the 6M drum with a locking-ring type closure subjected to a 4.9-foot drop. The drum is filled with water to 98 percent of overflow capacity. A three dimensional finite-element model consisting of metallic, liquid and rubber gasket components is used in the simulation. The water is represented by a hydrodynamic material model in which the material’s volume strength is determined by an equation of state. The explicit numerical method based on the theory of wave propagation is used to determine the combined structural response to the torque load for tightening the locking-ring closure and to the impact load due to the drop.

Corrosion Resistance of Iron-Based Amorphous Metal Coatings

New amorphous-metal thermal-spray coatings have been developed recently that may provide a viable coating option for spent nuclear fuel & high-level waste repositories [Pang et al. 2002; Shinimiya et al. 2005; Ponnambalam et al. 2004; Branagan et al. 2000–2004]. Some Fe-based amorphous-metal formulations have been found to have corrosion resistance comparable to that of high-performance alloys such as Ni-based Alloy C-22 [Farmer et al. 2004–2006]. These materials rely on Cr, Mo and W for enhanced corrosion resistance, while B is added to promote glass formation and Y is added to lower the critical cooling rate (CCR). Materials discussed in this paper include yttrium-containing SAM1651 with CCR ∼ 80 K/s and yttrium-free Formula 2C with CCR ∼ 600 K/s. While nickel-based Alloy C-22 and Type 316L stainless steel lose their resistance to corrosion during thermal spraying, Fe-based SAM1651 and Formula 2C amorphous-metal coatings can be applied with thermal spray processes without any significant loss of corrosion resistance. In the future, such corrosion-resistant thermal-spray coatings may enable the development of less expensive containers for spent nuclear fuel (SNF) and high-level waste (HLW), including enhanced multipurpose containers (MPCs), protected closure welds, and shields to protect containers from drips and falling rocks. These materials are extremely hard and provide enhanced resistance to abrasion and gouges from backfill operations. For example, Type 316L stainless steel has a hardness of approximately 150 VHN, Alloy C-22 has a hardness of approximately 250 VHN, while the Fe-based amorphous metals typically have hardness values of 1100–1300 VHN. Both Formula 2C and SAM1651 have high boron content which allow them to absorb neutrons, and therefore be used for enhanced criticality control. Cost savings can also be realized through the substitution of Fe-based alloy for Ni-based materials. Applications are also envisioned in oil & gas industry.

Transportable Laser Peening System for Field Applications to Improve Fatigue and SCC Resistance of Offshore Components and Structures

Laser peening technology has matured into a fully qualified production process that is now in routine and reliable use for a range of aerospace alloys. The technology is capable of extending the fatigue life and stress corrosion cracking life of components, and will enable designers to consider higher stress levels in life limited designs. Applications under development for steels include high and medium strength steels used in off shore oil exploration and production, titanium, aluminum and even ceramics and plastics as well as life extension of steel and aluminum welds. Fixed systems to treat components and transportable systems capable of field operations are available with a moveable beam that allows peening directly as needed on large structures.

Don’t Miss “Obvious” Hazards: Lessons Learned From Chemical Safety Board Investigations

The United States Chemical Safety & Hazard Investigation Board (CSB) was conceived by Congress following a series of catastrophic industrial accidents in the mid to late 1980s. This federal agency is charged with investigating incidents at chemical and manufacturing facilities, determining the causes, and making recommendations to prevent future accidents. This paper focuses on the findings from several CSB investigations related to equipment failure. Numerous codes, standards, and good practice guidelines are in place to govern the design, maintenance, and operation of vessels. However, the CSB has found that serious accidents continue to occur because of poor implementation of established guidance. This paper uses actual case studies to illustrate problems with equipment that ultimately led to catastrophes. Lessons learned from these incidents include designing equipment with adequate overpressure protection, adjusting inspection frequencies based on actual observations, and requiring written procedures for critical phases such as startup. Additional good practices and recommendations from the CSB are discussed with each of the case studies.

Justification for Onsite Transfer of Type B Packages After Extended Storage

This paper offers a practical means of qualifying previously loaded Type B packages for transportation onsite within the DOE complex after years of protected storage, while supporting the DOE program to maintain radworker dose as low as reasonable achievable (ALARA). Specifically, the paper discusses relevant packaging components and introduces part of a surveillance program carried out at the Savannah River Site supporting long-term storage of 3013-processed plutonium-bearing materials within closed 9975 packages and its application to DOE’s Equivalent Safety. Under normal service, maintenance is carried out annually to re-qualify the 9975 packagings for leak-tight transportation service. While in storage, however, annual maintenance was judged not to provide a significant increase in safety but to increase storage operation costs and to violate ALARA principles. Hence, a surveillance program was developed to investigate and confirm predictions of storage-related behavior for 9975 packaging materials, including the performance of O-ring seals and Celotex® insulation. The combination of analytical evaluations with surveillance data is shown sufficient to ensure that the 9975 packages can accommodate 1) time at storage temperature and 2) cumulative radiation dose without compromising subsequent performance under regulatory Normal Conditions of Transport or site-specific credible accident conditions.

Use of an Offset in Assessing Radiation Embrittlement Data and Predictive Methods

Analyses of reactor pressure vessel (RPV) surveillance data from Charpy V-notch shift results coupled with our latest knowledge of the mechanisms of radiation embrittlement have led to new predictive correlations/models that have a strong technical underpinning. In this paper we examine how well the new CRIEPI embrittlement predicts US RPV surveillance data. Secondly, we note that within the US surveillance data sets there are indications that the data may follow the same form as the predictive models, but the data may be offset by a constant amount (either positive or negative) from the predictive values. This offset can be attributed in some cases to inadequate baseline data. In other cases, there does not appear to be a constant offset, or such an offset is hidden by data scatter. This paper also reviews the potential use of an offset adjustment and focuses on several surveillance datasets for comparisons.

Fracture Toughness Characterization of Midland Beltline Weld After High Dose of Irradiation

The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory includes a task to investigate the shape of the fracture toughness master curve for reactor pressure vessel steel highly embrittled as a consequence of irradiation exposure, and to examine the ability of the Charpy 41-J shift to predict the fracture toughness shift. As part of this task, a low upper-shelf WF-70 weld obtained from the beltline region of the Midland Unit 1 reactor pressure vessel was characterized in terms of static initiation and Charpy impact toughness in the unirradiated and irradiated conditions. Irradiation of this weld was performed at the University of Michigan Ford Reactor at 288°C to neutron fluence of 3.4×1019 neutron/cm2 in the HSSI irradiation-anneal-reirradiation facility. This reusable facility allowed the irradiation of either virgin or previously irradiated material in a well-controlled temperature regime, including the ability to perform in-situ annealing. This was the last capsule irradiated in this facility before reactor shut down. Thus, the Midland beltline weld was irradiated within the HSSI Program to three fluences — 0.5×1019; 1.0×1019; and 3.4×1019 neutron/cm2. It was anticipated that it would provide an opportunity to address fracture toughness curve shape and Charpy 41-J shift compatibility issues at different levels of embrittlement, including the highest dose considered to be in the range of the current end of life fluence. It was found that the Charpy 41-J shift practically saturated after neutron fluence of 1.0×1019 neutron/cm2. The transition fracture toughness shift after 3.4×1019 neutron/cm2 was only slightly higher than that after 1.0×1019 neutron/cm2. In all cases, transition fracture toughness shifts were lower than predicted by the Regulatory Guide 1.99, Rev. 2 equation.

Fracture Toughness of Stainless Steel Cladding for Evaluation of the Degraded Davis-Besse RPV Head

Boric acid attack in the reactor pressure vessel (RPV) head of the Davis-Besse (D-B) nuclear plant led to wastage through the 150-mm low alloy steel head such that the stainless steel cladding was exposed. The Heavy-Section Steel Technology (HSST) Program at Oak Ridge National Laboratory was commissioned by the Nuclear Regulatory Commission to conduct a program of testing and analysis to enable an evaluation of the structural significance of cladding defects found in the wastage cavity of the D-B head. The overall test program consisted of material characterization at 316°C (600°F) of cladding materials, pressure vessel burst tests of cladding discs with and without flaws, and extensive analytical studies. Three different cladding materials were tested and evaluated, one from an unused commercial RPV that was used for the clad-burst experiments, an archival cladding previously used for various experimental and irradiation experiments, and the cladding from the D-B head. This paper compares and discusses the fracture toughness test results conducted with the three claddings, and the fractographic analyses conducted on the clad-burst discs. Comparison of J-resistance curves for the three clad materials shows significant material variability and disparity in the results from two test specimen types. Fractographic examinations of clad-burst discs showed transition from ductile tearing to shear mode of fracture. The relationship of the cladding test results with the clad-burst results is discussed.