Control Rod Drive Research Articles

BWR lower head penetration failure test focusing on eutectic melting

Decommissioning work occasioned by the Fukushima Daiichi Nuclear Power Station (1F) accident of March 2011 is in progress. Severe accident (SA) analysis, testing, and internal investigation are being used to grasp the 1F internal state. A PWR system that refers to the TMI-2 accident is typical for SA codes and testing, on the other hand, a BWR system like 1F is uncommon, understanding the 1F internal state is challenging. The present study conducted the ELSA-1 test, a test that focused on damage from eutectic melting of the liquid metal pool and control rod drive (CRD), to elucidate the lower head (LH) failure mechanism in the 1F accident. The results demonstrated that depending on the condition of the melt pool formed in the lower plenum, a factor of LH boundary failure was due to eutectic melting. In addition, the state related to the CRD structure of 1F unit 2 were estimated.

Verification of neutronics and thermal-hydraulics coupling method for FLUENT code using the MSRE pump startup, trip data

The major objective of the present study is to verify a method for the FLUENT code applying the user defined function (UDF) which enables the coupling simulation of neutronics and thermal-hydraulics of a molten salt fast reactor based on the one-point kinetics. The UDF with the added logic to control the control rod drive (CRD) is linked to the FLUENT code to analyze the reactivity change when the fuel pump of the molten salt reactor MSRE was started and tripped. These experiments were carried out under zero power conditions. The core of the MSRE is modeled on a cylinder with the same flow channel area as MSRE, and the control rod in the simulation model can compensate for the reactivity changes that accompany with the flow rate changes. When the pump is started, the response of the control rod overshoots because the reactivity change is in a form close to the stepwise change. This behavior has been reproduced almost correctly using a simple proportional-integral-derivative (PID) controller program. The response when the control constant is changed has been also investigated. In the fuel pump trip experiment, the transient change is relatively moderate. This behavior is also reproduced almost correctly with the same control logic. These simulations show that the UDF has been programmed correctly, and the discretized point kinetic model with prompt jump assumption can be applied to the molten salt reactor in combination with the code-to-code benchmarks already performed.

MAAP 5.04 ex-vessel CRD tube heat transfer and melting model

Enhancements are made in the Boiling Water Reactor (BWR) version of MAAP 5.04 to consider ex-vessel Control Rod Drive (CRD) housing heat transfer and melting. The in-vessel CRD tube model is extended below the reactor vessel and conduction and radiative heat transfer between the lower head, CRD tubes, pedestal gas and pedestal wall are considered. The possibility of corium melt flowing into collapsed CRD tubes and causing creep rupture of CRD tubes outside the reactor vessel, leading to vessel failure, is considered. Lastly, ex-vessel CRD tube melting due to the corium jet exiting the breached reactor vessel after vessel failure is considered. The BWR Station Blackout (SBO) scenario was simulated using the new code. The results show that explicitly modeling the ex-vessel CRD tube heat transfer did not affect the time of vessel failure. Also, vessel failure due to corium ingress into ex-vessel CRD tubes was not predicted. A small, but nontrivial mass of steel from ablated ex-vessel CRD tubes was added to the debris bed on the pedestal floor after vessel failure.

Analysis of dust and fission products in a pebble bed NGNP

In the HTR pebble bed reactor graphite dust is generated during normal reactor operation due to pebble-to-pebble interactions. This dust will be deposited throughout the primary system. Furthermore, the dust will become radioactive due to sorption of fission products released during normal operation.This paper presents an analysis of dust transport and deposition and fission product transport and plate-out during normal operation of a NGNP (next generation nuclear plant) Westinghouse pebble bed design. The main objective is to determine the amount and location of deposited graphite dust in the system and the amount of radioactive isotopes adsorbed on the structures and the dust during normal operation. The results will be used in planning of maintenance activities. Moreover the present results may be used in a next step to perform a depressurized loss of forced cooling (D-LOFC) analysis and to determine the amount of radioactivity released to the atmosphere during the accident. The analysis was performed using the SPECTRA code.Based on the assumed dust source term, during the 60 years lifetime of the pebble bed reactor concept which was analyzed approximately 1630kg dust enters the primary helium flow. It was determined:(1)That 86% of the dust settles on the graphite structures inside the reactor vessel. Of the remaining graphite dust 2/3 collects in the low temperature intermediate heat exchanger (IHX) and the remaining dust in the high temperature IHX and the connecting pipes of the primary system.(2)Agglomeration of dust from smaller into larger particles is observed at all locations. The resuspension of agglomerated dust particles is an important phenomenon that limits the build-up of a dust layer on the surface of the IHXs. A large deposited dust layer is observed on the control rod drive (CRD) walls in the graphite structures. The dust accumulates on the CRD walls because, due to very low gas velocities, resuspension of the agglomerated particles does not occur.(3)The dust binds a significant amount of fission product vapors. Although the vapor concentrations are typically smaller on the dust surface than on the metallic surfaces, the dust particles have a relatively large surface.

Precise Evaluation of Corrosion Environments of Structural Materials under Complex Water Flow Condition, (I)

To estimate the corrosion potentials of structural materials under complex water flow condition, such as reactor pressure vessel (RPV) bottom region of Boiling Water Reactors (BWRs), a method was newly developed. Three-dimensional water flow analysis was performed and corrosion potentials of structural materials were calculated on the basis of the flow analysis results. Water flow analysis showed the velocities in the RPV bottom region varied from about 0.1 ms-1 to 4.5 ms-1. From the corrosion potential estimation under hydrogen water chemistry (HWC) condition, at the jet pump outlet and shroud support leg, a rather large amount of hydrogen had to be added to reduce the potential because of high flow velocity condition. Conversely, a small amount of hydrogen was sufficient in the case of the stub tube of the control rod drive (CRD) guide tubing located in the center of the bottom region.

A simplified model of the BWR depressurization transient

A simplified model is proposed to evaluate the BWR plant depressurization transient subsequent to the activation of a single safety relief valve. The model is validated with a series of RELAP5 Mod 3 calculations yielding reasonable comparisons. The depressurization behavior of the 26 US BWR plants is analyzed with the model and the relevant parameters controlling the process are identified. Of these, four are plant-specific and their effect on the final outcome of the depressurization transient is quantified. The minimum vessel inventory at the activation of the low pressure emergency core coolant injection is selected as the figure of merit. Core power is the only parameter that significantly affects the outcome of the transient. The role of the low pressure injection and of the discharge port cross-sectional area is discussed in detail. The effect of operator actions on the control rod drive (CRD) injection flow is also discussed.

BWR control rod drive system aging

Abstract This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control rod drive mechanisms (CRDMs) and assesses the merits of various methods of “managing” this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each U.S. BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components affected by the effects of service wear and aging are valve seals, discs, seats, stems, packing, and diaphragms.

Detecting And Mitigating Rod Drive Control System Degradation In Westinghouse PWR's

A study of the effects of aging on the Westinghouse Control Rod Drive (CRD) System was performed as part of the U.S. NRC's Nuclear Plant Aging Research (NPAR) Program [l], [2]. For the study, the CRD system boundary includes the power and logic cabinets associated with the manual control rod movement, and the control rod mechanism itself. The aging-related degradation of the interconnecting cables and connectors and the rod-position-indicating system also were considered. This paper presents the results of that study pertaining to the electrical and instrumentation portions of the CRD system, including ways to detect and mitigate system degradation.

Detecting and mitigating rod drive control system degradation in Westinghouse PWRs

A study of the effects of aging on the Westinghouse Control Rod Drive (CRD) System was performed as part of the US NRC's Nuclear Plant Aging Research (NPAR) Program. For the study, the CRD system boundary includes the power and logic cabinets associated with the manual control rod movement, and the control rod mechanism itself. The aging-related degradation of the interconnecting cables and connectors and the rod-position-indicating system also were considered. The authors present the results of that study pertaining to the electrical and instrumentation portions of the CRD system, including ways to detect and mitigate system degradation.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Ultrasonic examination of austenitic welds at reactor pressure vessels

Austenitic welds play an important role in reactor pressure vessels (RPV) both of pressurized water reactors (PWRs) and boiling water reactors (BWRs). These pressure retaining joints still pose problems for ultrasonic examination. During the last few years some progress has been achieved in overcoming these problems. The state of the art of the practical application is described in “The Handbook on Ultrasonic Examination of Austenitic Welds” of the International Institute of Welding which was issued in 1986. This document was prepared by an international working group. Sulzer Brothers Limited, Winterthur Switzerland as manufacturer of both BWR and PWR RPVs has a long experience with ultrasonic examination of austenitic welds. In General Electric BWR RPVs the main recirculation line, the feedwater line and additional lines are connected to nozzles of the RPV with dissimilar metal welds joining stainless steel to carbon steel. The control rod drive (CRD) housings are attached to the RPV with complex weld joints. Framatome and Westinghouse RPVs also are provided with welds between the vessel and the main coolant line. This paper presents examples of ultrasonic approaches and examination results. Aspects of fabrication, preservice and inservice inspections are also covered. Complex weld joint designs both from the point of material and geometry make ultrasonic examination sometimes extremely difficult but based on case by case investigations reliable solutions can often be realized.

Automatic Refueling Platform System and Automatic Control Rod Drive Remote Handling System for BWR Nuclear Power Plants

This paper introduces some of the mechanized and automated systems already applied to nuclear power plants in service, and describes the resulting significant contributions to shorter outage time, reduced personnel exposure, and improved plant reliability. The systems discussed are an automatic refueling system and an automatic remote handling system for control rod drive (CRD).