水冷反应堆燃料元件的在役检测和处理

水冷反应堆包括轻水堆和重水堆,轻水堆分为压水堆和沸水堆;重水堆分为加压重水堆和加拿大的氘铀堆。国际上把它们归为一类进行研究。本文涉及的破损燃料元件的在役检测和处理包括：反应堆运行时的检测;换料时或换料后的检测;在燃料组件内鉴别破损的燃料棒;燃料组件的监测、拆卸和修复;破损燃料棒拆出后的检测,破损定位与修补。     

Development of Pressure-Tube Inspection Equipment and Experience in “FUGEN” (ATR)

Abstract

Remote-controlled in-service inspection (IS1) equipment has been developed for inspecting the pressure tubes (117.8mm I.D., 4.3mm thickness, 4.8m length, made of Zr alloy) of the Fugen, which is a heavy-water-moderated, boiling-light-water-cooled pressure-tube-type reactor. The equipment is capable of performing three kinds of inspection; ultrasonic flaw detection, measurement of inside diameter and cross-sectional form, and visual inspection of the internal surface.

A system mock-up was prepared using facilities outside the reactor; the test results were:

(1) Detection of a flaw of 5.0 mm (L) x0.06 mm (W) ×0.05 mm (D) with S/N=15 dB

(2) Inside diameter measured to ±15 µm for water temperatures of 10–407°C

(3) Identification of a 2.0 mm (L) x0.1 mm (W) ×0.1 mm (D) flaw

(4) Under a radiation level of 3×10⁵ R/h (Gamma), the system was confirmed to be reliable for 80 h of operation.

This equipment was used for inspecting of 15 pressure tubes in September 1989 during the 8th annual inspection of Fugen, and the inspection was finished on schedule.     

水冷反应堆燃料元件的在役检测和处理

水冷反应堆包括轻水堆和重水堆,轻水堆分为压水堆和沸水堆;重水堆分为加压重水堆和加拿大的氘铀堆。国际上把它们归为一类进行研究。本文涉及的破损燃料元件的在役检测和处理包括：反应堆运行时的检测;换料时或换料后的检测;在燃料组件内鉴别破损的燃料棒;燃料组件的监测、拆卸和修复;破损燃料棒拆出后的检测,破损定位与修补。     

Piping Cracks in JPDR, (II)

Ultrasonic and radiographic examinations of weldments in the primary piping of the Japan Power Demonstration Reactor (JPDR) revealed crack indications in two reactor vessel nozzle safe-end to pipe joints CS-1 and UL-1, while on the joint FW-132 a crack-like indication was given only on the ultrasonic record.

These three parts of the reactor piping were removed from the reactor, and subjected to thorough nondestructive and destructive examination in laboratory.

The indications obtained in the field by ultrasonic and radiographic means agreed with each other, and were further confirmed by the laboratory tests.

Crack indications given by one nondestructive method and not concurred by another method may safely be taken as inconsequential on spurious.     

Development of Multi-Functional Telerobotic Systems for Reactor Dismantlement

This report summarizes technological features of advanced telerobotic systems for reactor dismantling application developed at the Japan Atomic Energy Research Institute. Taking into consideration the special environmental conditions in reactor dismantling, major effort was made to develop multifunctional telerobotic system of high reliability which can be used to perform various complex tasks in an unstructured environment and operated in an easy and flexible manner.

The system development was carried out through constructing three systems in succession; a light-duty and a heavy-duty system as a prototype system for engineering test in cold environment, and a demonstration system for practical on-site application to dismantling highly radioactive reactor internals of an experimental boiling water reactor JPDR (Japan Power Demonstration Reactor). Each system was equipped with one or two amphibious manipulators which can be operated in either a push-button manual, a bilateral master-slave, a teach-and- playback or a programmed control mode. Different scheme was adopted in each system at designing the manipulator, transporter and man-machine interface so as to compare their advantages and disadvantages.

According to the JPDR decommissioning program, the demonstration system was successfully operated to dismantle a portion of the radioactive reactor internals of the JPDR, which used underwater plasma arc cutting method and proved the usefulness of the multi-functional telerobotic system for reducing the occupational hazards and enhancing the work efficiency in the course of dismantling highly radioactive reactor components.     

Inspection techniques for primary pressurized water cooler tubes in the high temperature engineering test reactor

A primary pressurized water cooler (PPWC) with 136 inverse-U-tubes is installed in the primary cooling system of the high temperature engineering test reactor (HTTR). The HTTR is the first high temperature gas-cooled reactor in Japan with an outlet gas temperature of 950 °C and thermal power of 30 MW. The heat transfer tubes form the reactor pressure boundary of the primary coolant. Inspection techniques for the tubes should be established to carry out the in-service inspection efficiently. An automatic inspection system for the tubes uses probes for eddy current testing and ultrasonic testing. Defect detecting characteristics of the inspection probes and the application of the automatic inspection system to nondestructive test of the tubes were examined by a mockup test utilizing artificially degraded tubes. The automatic inspection system could smoothly insert and withdraw the probe at its moving velocity in the fixed positions of the defected tube. Nondestructive test of the tubes using the automatic inspection system was conducted during reactor shutdown period of the HTTR after test operation of about 55% of the full power. Through the nondestructive test, it was concluded that there was no defect on the outer surface of the heat transfer tubes of the PPWC inspected.     

The Development of an Apparatus for Providing Artificial Surface Defects on Inside Surface of Cladding Tubes

It is important from the viewpoint of reactor safety as well as from economical interest to determine the limit of allowable depth and length of surface defects on fuel cladding tubes. For the purpose of preparing cladding tubes provided with artificial defects of various degrees to be used as standard reference specimens for internal pressure burst tests, an apparatus has been developed for plowing troughs of predetermined depth and length on the inside surface of cladding tubes.

This apparatus is capable of cutting such troughs in the center of a cladding tube 100 cm long, and the depth of the scratched defect can be measured nondestructively with the use of a measuring instrument provided with stylus.

The paper presents an outline of the apparatus together with some experimental results.     

Techniques for processing remote field eddy current signals from bend regions of steam generator tubes of prototype fast breeder reactor

Steam generator (SG) is one of the most critical components of sodium cooled fast breeder reactor. Remote field eddy current (RFEC) technique has been chosen for in-service inspection (ISI) of these ferromagnetic SG tubes made of modified 9Cr–1Mo steel (Grade 91). Expansion bends are provided in the SGs to accommodate differential thermal expansion. During ISI using RFEC technique, in expansion bend regions, exciter–receiver coil misalignment, bending stresses, probe wobble and magnetic permeability variations produce disturbing noise hindering detection of defects. Fourier filtering, cross-correlation and wavelet transform techniques have been studied for noise reduction as well as enhancement of RFEC signals of defects in bend regions, having machined grooves and localized defects. Performance of these three techniques has been compared using signal-to-noise ratio (SNR). Fourier filtering technique has shown better performance for noise reduction while cross-correlation technique has resulted in significant enhancement of signals. Wavelet transform technique has shown the combined capability of noise reduction and signal enhancement and resulted in unambiguous detection of 10% of wall loss grooves and localized defects in the bend regions with SNR better than 7 dB.     

Development of Telerobotic Systems for Reactor Decommissioning, (III)

This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program.     

Derivation of BWR Core Dynamics Models and Analysis of JPDR Core Transient Tests

The purpose of the present work is to establish a core dynamics model for the JPDR plant (natural circulation boiling water reactor plant). First, upon surveying the analytical model with the so-called distributed parameters in reference to the experimental results obtained during the JPDR initial power-up test, it was decided to develop the core dynamics model with lumped rather than distributed parameters, and to further modify it to obtain the highest possible accuracy. Three equations for mass, energy and momentum balances were used, and particular care was taken in the determination of the core void fraction and recirculation flow, the development of the thermo-hydrodynamic equation in the saturated region, and the choice of void sweep time as well various numerical constants.

The derived dynamics model was then applied to the analysis of the JPDR transient test involving Bypass Regulator and Initial Pressure Regulator operations, as well as to that of the Bypass Regulator oscillation test.

The calculated results by this model agreed well with data from experiment.

A comparison was then made between the model and other core dynamics models. It was concluded that this model intrinsically has good accuracy and is easy to handle; it should prove very useful for core dynamics analysis of any BWR, in their relation to the out-of-core controlling system.     

Inservice inspection of the reactor block of sodium-cooled fast breeder reactors

By using sodium as coolant special boundary conditions result for the inservice inspection (ISI) of fast breeder reactors. For that reason in general it is not successful applying the methods and equipment proved for the 151 of light water reactors.This report presents inspection methods and equipment developed for the ISI of the reactor block of sodium cooled fast breeder reactors. The survey takes into account the state of the art as well as some R&D-work at home and abroad. Entering into particulars the methods and equipment used for leak monitoring, the inspection of the reactor vessel wall, the inspection' of reactor internals above and below the sodium level, monitoring of structure home noise and the measurement of the gap between the reactor vessel and the guard vessel are described.     

Sensitivity analysis of input parameters for pressure tube integrity evaluation

Pressure tubes in CANDU reactor are the most important components that contain fuel. The operating experiences show leaks and burst in pressure tubes over the past two decades. The integrity of pressure tubes is, therefore, key concern in CANDU reactor. Once a CANDU reactor put into operation, their integrity could be checked by in-service-inspection. However, comparing to the total number of pressure tubes in a CANDU reactor, only a small number of pressure tubes are selected for inspection, since there is no weld. The inspection scope and results have been treated so far using a deterministic approach. Taking into account the difficulty in inspection sampling and in extrapolating the results to the entire core, a probabilistic approach is necessary. In this study, probabilistic integrity assessments are carried out considering key factors, such as initial hydrogen concentration, defect shape, delayed hydride cracking (DHC) velocity and fracture toughness. The leak and failure probabilities are calculated as a function of time by applying Monte Carlo simulation.     

PISC full scale reactor pressure vessel validation of non-destructive examination

The PISC III Programme involves validation of techniques and procedures and, within this programme, evaluation has now started on the ability to discriminate service induced defects from indications produced by fabrication defects in A 508 Class 2 material when sensitive techniques are used.Action No. 2 of PISC III: Full Scale Vessel Testing is designed for the performance demonstration of three groups of inspection procedures:

• - Mechanized ASME type procedures with variable recording level and complementary techniques

• - Industrial full ISI procedures (mechanized);

• - Several detailed evaluation procedures (generally mechanized) based on advanced techniques to be used on defective areas detected by usual inspection.

These procedures, typical for ISI in most of the cases, are applied in four situations which could be typical of old and new LWR pressure vessels:

• - vessel material and welds containing important service and fabrication defects but mixed with base material defects and small welding defects;

• - nozzle to shell welds with typical service defects, often well isolated and distant from other defective areas in rather clear material and/or welds;

• - nozzle inner radius defects;

• - artificially heat and unbranched fatigue defects in the test blocks assembled to simulate a PWR pressure vessel wall portion.

The paper summarizes the PISC II programme results which stress the characteristics of capable NDT techniques, in opposition to material characteristics like acceptable base material defects. It describes the full scale pressure vessel components available to conduct the PISC III exercise with improved ultrasonic techniques.     

Pressure vessel integrity in the presence of defects: The probabilistic prediction in spite of scarce data

A closed-form solution is presented for the probability of fast fracture of a pressure vessel in the presence of crack-like defects. The defect distributions are characterized by the total number of defects and the concentrations and lower threshold values of exponential size distributions. The effect of ultrasonic inspection on defect occurrence is taken into account. The structure of the failure probability is examined, and it is shown what types of defect distributions may be admitted in order to keep the likelihood of failure reasonably remote. Remarkably specific conclusions can be drawn given the limited data available.     

Risk-informed inservice inspection program

The Nuclear Regulatory Commission (NRC) has developed draft guidance for power reactor licenses on acceptable methods for using probabilistic risk assessment (PRA) information and insights in support of plant-specific applications to change the current licensing basis (CLB) for inservice inspection (ISI) of piping. This process is also known as risk-informed inservice inspection programs (RI-ISI). The risk-informed inservice inspection process for operating nuclear power plants provides an alternative method for selecting and categorizing piping components that are inspected for the purposes of meeting the requirements of ASME Section XI. A RI-ISI approach will incorporate probabilistic techniques to help define the scope, type and frequency of inservice inspection. The risk-informed process may support a decrease in the number of inspection and inspection intervals but will also identify areas where increased resources should be allocated to enhance safety. The approach discussed in this paper follows the method developed by NRC staff.     

Pressure vessel integrity and weld inspection procedure

The primary objective of this study is to develop a simple methodology which, when coupled with existing observations on pressure vessel behavior, provides an interrelation between nuclear pressure vessel weld integrity and the parameters of the in-service inspection program, including inspection sample size, frequency and efficiency. The basic input information on rate of generation and development of weld flaws of different sizes and types is drawn primarily from published British and German studies taken almost exclusively from welds of non-nuclear pressure vessels. The input information is varied to reflect differences in weld quality and uncertainty of input data. A modified Markov process is employed and a computer code written to obtain numerical results. If it is assumed that the quality of nuclear reactor welds are the same as the quality of non-nuclear welds (i.e. the data base), then, based on the limitations of the model, the predicted critically sized defect concentration is about 50 × 10⁻⁷ per weld at the end of weld life for welds under both high and low stress if ASME, Section XI, In-Service Inspection Requirements are applied. Based on the British data and the less stringent inspection standards (compared to Section XI) the estimated number of critically sized defects per weld at the end of weld life is 250 × 10⁻⁷ and 170 × 10⁻⁷ per weld for high and low stressed welds, respectively. If it is assumed that the nuclear reactor pressure vessel welds have superior quality to the non-nuclear welds, then the model predicts an appropriately lower probability of critical defects at the end of weld life. A variety of other sensitivity studies are included in the report. Also, a simple methodology to provide an optimal weld inspection program which is consistent with a minimum cost criteria is outlined. It should be noted that the results of this study are based on the limitations of the simple model that was used and on a variety of corresponding assumptions.     

压水堆核电厂核岛机械设备在役检查规则研究

压水堆核电厂核岛机械设备在役检查规则研究是修订和编制我国相关核电在役检查标准的基础和前提。本文简介了在役检查规则研究目标、方法、主要内容和结果以及在役检查规则制定依据,简述了规则研究相关主要问题的处理方法和结果,对比分析了依据研究结果编制的NB/T 20312标准与EJ/T 1041标准在役检查规则的主要不同点,给出了准确理解和正确应用NB/T 20312标准有关在役检查规则的提示和说明,为有效应用该标准在役检查规则提供重要参考。     

Requirements related to in-service inspection of reactor vessels

The in-service inspection of nuclear facilities should make it possible to detect all flaws capable of deterioration. The size of a flaw is not normally characteristic of its seriousness: a large, stable defect which is not developing is of no importance for the safety of the facility whereas a small but developing defect may be dangerous and hence of importance for safety. It is essential to detect all defects, so that any development they may be undergoing can be assessed, and to use, for that purpose, a technique with a maximum detection probability; but the main thing is to characterize all developing defects. For the in-service inspection of pressurized-water reactor vessels the “Commissariat à l'Energie Atomique” has conducted studies on inspection methods with a view to ensuring first the best possible detection and secondly the best possible characterization of flaws. Since 1973 many publications have appeared on the subject and this paper will only recall the essentials giving particular attention however to the characterization of defects, which is the principal purpose of the studies currently being conducted.     

Steam generator small bore piping socket weld inspection using the phased array ultrasonic technique

As nuclear power plants age, the likelihood of failures in the small bore piping used in those plants caused by exposure to mechanical vibrations during plant operations increases. While small bore piping failures rarely cause plant shutdown, the management of small piping has been a keen area of interest since their repair or maintenance requires a reactor trip. Steam generator (SG) drain pipe socket welds are small diameter piping connections (nominal pipe schedule 3–4 inches) susceptible to mechanical vibration. SG drain pipe socket weld failures have caused coolant leakage. Therefore, more reliable inspection technologies for small bore piping need to be developed to detect problems at an early stage and prevent pipe failures. This research aims to improve the reliability and accuracy of small bore piping inspections through the design, manufacture and application of a new phased array ultrasonic testing technique and inspection system for SG drain line socket welds.     

Residence Time of Crud on Surfaces of Channel Box in JPDR

Characterization of crud on surfaces of the channel box in JPDR has been carried out by means of chemical, radiochemical, X-ray diffraction and infrared spectrum analyses. The main cations in the crud are Fe and Ni: The sum of their weights amounts to more than 90% of the total weight of the cations found. The results of X-ray diffraction and infrared spectrum analyses revealed that the crud consisted of Ni_0.65Fe_2.35O₄, NiO and γ-FeOOH.

Based on the neutron flux calculated from the burn-up of ²³⁵U and ²³⁸U in the spent fuel, the apparent residence time of elements on the surface of the channel box was calculated to be 230 d for Co, 260 d for Ni and 70 d for Fe. The value for Fe should be taken as a minimum value, because of the presence of γ-FeOOH in the crud, which has been formed during the storage in a pond.

The present data are discussed in correlation with the one in the reactor water.