Radiation dose management in nuclear power plants

Pergamon

www.elsevier.com/locate/pnucene

Progress in Nuclear Energy, Vol. 39, No. 1, pp. 31-51, 2001 0 2001 Published by Elsevier Science Ltd. All rights reserved

Printed in Great Britain 0149-1970/01/$~ see front matter

PII: SO149-1970(01)00002-6

Radiation Dose Management in Nuclear Power Plants

Man-Sung Yim North Carolina State University

Raleigh, NC 27695

Howard Ocken Electric Power Research Institute

Palo Alto, CA 94304

ABSTRACT

Various approaches available for radiation dose management in nuclear power plants are reviewed in this paper. These approaches include radiation field reduction, accurate dose reporting, good health physics planning, exposure control during job execution, and implementation of a radiation protection culture. For the reduction of radiation fields, a wide variety of technological methods are also available. These include controlling cobalt sources, preconditioning of replacement components, controlling primary chemistry, controlling reactor shutdown and startup chemistry, chemical decontamination, minimizing fuel failures, and using sub-micron purification filters. Successful radiation management requires cooperation among the various groups with relevant responsibilities at a nuclear power plant if the desired exposure goals are to be achieved. 0 2001 Published

by Elsevier Science Ltd. All rights reserved.

1. Introduction

The goal of radiation protection is to reduce the probability of radiation-induced diseases in persons exposed to radiation and any associated genetic effects. Dose reduction measures are to be implemented to the degree that the expected benefits from such activities can be realized at a given plant. To achieve this goal, various radiation protection guidelines and regulatory standards have been developed. These have undergone changes over the last several decades, reflecting recent development in radiation biology. The latest International Commission on Radiological Protection (ICRP) recommendations in its Publication 60 [ICRP, 19911, hereafter called ICRP-60, reflect the revised risk coefficient relating to exposure to ionizing radiation that became available in the second half of the 1980s. ICRP-60 also notes that stochastic deleterious health effects of radiation cannot be completely avoided because of the difficulties in invoking a threshold for them.

The ICRP-60 has led to significant changes in attitudes and to a new culture of radiation dose management in nuclear power plants. The primary dose limits for workers are an effective dose of 20 mSv per year averaged over five years with not more than 50 mSv in any single year. There is also considerable emphasis on the need to optimize protection within the constraints. Lower dose limits and the requirement to ensure that exposure is as low as reasonably achievable (ALARA) means that detailed attention must be given to radiation dose management in the design, operation, and decommissioning of nuclear power plants.

31

32 M.-S. Yim nnd H. O&en

In many countries, ALARA was a major part of radiation protection programs in nuclear power plants prior to ICRP-60. Efforts in ALARA resulted in a significant occupational dose reduction in these countries [Wilson, et al., 1986; Khan and Baum, 1988; Ocken, 1997; NCRP, 19941. In the U.S., between 1983 and 1996, a five-fold decrease in radiation dose was observed in nuclear power plants [Ocken, 19971. After 1996, the dose continued to decrease but at a slower rate [Wood, 1999a]. In the countries where the historical records of radiation exposure were low, however, the opposite trend of increased radiation dose has been noticed [NCRP, 19941. This indicates the continued importance of ALARA application [Thezee, et al., 19951. ICRP-60 provides additional incentives for continued and further enhancement of ALARA in plant radiation protection.

The purpose of this paper is to review various approaches to the development of effective schemes for radiation dose management in an operating nuclear power plant reflecting recent technological developments.

2. Overview of Radiation Dose Management

The underlying theme of radiation dose management in nuclear power plants is to protect workers from radiation risks. ICRP-60 recommendations as the latest development in basic principles of radiation protection provide justification for both the use of individual limits and the optimization of radiological protection.

ICRP-60 notes misconceptions in practice as to the definition and function of dose limits - sometimes dose limits are mistakenly seen as the boundary between that which is safe and that which is dangerous [Boehler, 199.51. This concept of limit is neither an assurance of quality radiological protection nor should it be considered the ultimate purpose of radiation protection. According to ICRP-60, what needs to be in place is a system of radiological risk management based on the “prudent avoidance” principle with the recognition of stochastic health effects of radiation. ICRP-60 stresses not only compliance with the principle of dose limitation but also a new logical, rational and effective approach to protection based on the quest for a balance between the cost of protection and levels of residual individual and collective exposure.

ICRP-60 states that both doses and risks should be optimized within the dose and risk limits laid down for individuals. Therefore the limitation is acting as an individual guarantee of protection only where some individuals are exposed to excessive doses. The optimization approach provides protection at the individual and collective levels in so far as it provides optimal protection for all individuals. The aim of optimization is not only to reduce exposure to levels ALARA, taking economic and social constraints into account, but also to give priority to reducing the exposure of those with the highest levels of individual doses.

With new ICRP-60 recommendations, what is needed is the continued encouragement of a dynamic management approach to radiation protection in contrast to the strict application of individual dose limits. A management approach integrating various aspects of radiation protection with regards to the sources of exposure, the variety of jobs, and the different means of controlling or reducing radiation exposures based on the optimization of individual and collective doses should be in place. In the U.S., the ability to control

Radiation dose management 33

plant occupational radiation dose has been monitored by the Nuclear Regulatory Commission (NRC) as a tool to assess the effectiveness of utility management control of plant activities [Ocken, 19971.

For effective application of ALARA, priorities should be given to the high-dose jobs, and the effectiveness of any control measure should be judged by a cost/benefit comparison. The parameters that could be considered in this comparison include [NRCP, 19941: ( 1) work-hours saved (or increased), (2) dose avoided (or increased), (3) monetary value of dose reduction for the specified job (dollar per person-mSv), (4) capital cost, (5) engineering, design and analysis costs, (6) operations and maintenance cost, (7) power replacement costs, (8) current interest and discount rates, (9) depreciation period, (IO) any applicable taxes, (11) salvage value, (12) reliability, (13) waste disposal cost, (14) extension (or early termination) of existing plant license, and (15) decommissioning costs.

Typically high-dose jobs are known to include steam generator manway cover removal and replacement, eddy current testing and repairs of steam generator tubing, reactor refueling operations, and maintenance on reactor coolant pumps and valves [Baum and Matthews, 19851. Based on the US nuclear industry’s experience in the 80s details on the repetitive high-dose jobs in nuclear power plants were identified both for pressurized water reactors (PWRs) and boiling water reactors (BWRs) as shown in Table 1 [Dionne and Baum, 19851.

Table 1. Repetitive high-dose jobs in PWRs and BWRs [Dionne and Baum, 19851 PWR r 1 RWRc I ..I\Y Y ITI.

Cavity decontamination Control-rod drive removal/rebuild and replacement CVCS* repair and maintenance Fuel shuffle, leak detection and inspections Fuel-shuffle, leak detection and inspections In-service inspections (ISI) In-service inspections (ISI) Instrumentation repair and calibration Instrument repair and calibration Jet pump inspection and repair Insulation removal and replacement Main steam isolation valve repair and inspection Operations-surveillance, routines and valve lineups Operations-surveillance, routines and valve lineups Plant decontamination Plant decontamination Pressurizer valve inspection, testing and repair Primary valve maintenance and repair Primary valve maintenance and repair Radwaste system repair, operation and maintenance Radwaste system repair, operation and maintenance Reactor assembly/disassembly Reactor assemble and disassemble Recirculation pump seal replacement Reactor coolant pump seal replacement Refueling pool decontamination Residual heat removal system repair and maintenance Residual heat removal system repair and maintenance Scaffold installation/removal Safety valve repair and inspection Secondary side of steam generator inspection and repair Scaffold installation/removal Snubbers, hangers and anchor bolts inspection and repair Snubber inspection and repair Steam generator eddy current testing Transversing in-core probe calibration, repair and Steam generator manway removal/replacement maintenance

steam generator tube plugging/sleeving Torus repair, inspection and modification

* CVCS: chemical and volume control system Turbine overhaul and repair

34 M.-S. Yim and H. Ocken

Table 2 compares costs/benefits for different management options based on the US industry experiences in the 80s [Baum and Matthews, 19851. Although the information does not reflect the latest technological development, it provides some good examples. As seen in the Table, in addition to the use of proper shielding or facilitation of maintenance activities, reducing radiation sources provides sound benefits. Reducing radiation fields facilitates the regularly scheduled plant maintenance activities by preventing the adverse impact of any tightening of individual exposure limits. These adverse impacts include [Ocken, 19971: (1) unavailability of experienced workers for high dose jobs; (2) restricting exposure limits of contractor’s key workers; (3) high dose work may require use of more crews to limit individual exposures, and; (4) reduced flexibility for job planning, increased number of workers and health physics coverage, and extended outage time. Improved worker efficiency provides obvious financial benefits in addition to dose savings.

Various approaches are available for the reduction of radiation fields in operating plants [Ocken and Wood, 1992; Ocken, 1997; Bergmann and Bencini, 19971. These include exposure source reductions through material replacement, primary chemistry control, cobalt substitution, decontamination, and preconditioning of materials.

Advancement of information technology and computers also allows dramatic changes in the management of occupational dose in nuclear power plants [Bennett, 19961. Instead of relying on manually kept records of film badge dose information, use of electronic dosimetry with computerized databases enables rapid and effective dose control. Application of remote dose-rate measuring instruments allows continuous monitoring of the workplace. Real-time dose management of workers in the radiation area is possible by using radio-frequency technology and transmitting data from local radiation protection monitoring points to the main control area. Dose and time data for individuals and jobs can be recorded on a common basis by using a centralized database with data transfer links to the various sites. Doses for future work can be predicted by using computer based recording systems and data analyses. A predictive computer code capable of estimating dose-rates, doses and shielding requirements for tasks where historical data is not applicable can also be used. These allow effective planning for manpower requirements, identification of areas requiring dose reduction effort, and monitoring the effects of any changes in work practice or plant modifications. The effectiveness of this approach is guaranteed, because the careful planning of worker radiation protection and the implementation of guidelines have been demonstrated as having significant impact on dose reduction [Wood, 1995a].

The Information System on Occupational Exposure (ISOE) represents an international effort to support these efforts [Robinson and Lazo, 19951. ISOE is an international occupational exposure database and communications network for dose optimization. The ISOE databases include: (1) annual operational dosimetry data; (2) plant configuration and administrative data, and (3) details on the dosimetric results of specific operations. By properly planning, preparing, implementing, and reviewing jobs using these databases, significant dose reductions could be made. More details of dose management approaches are described in the next section.

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4. Methods of Radiation Dose Management

The methods of dose management can be categorized into: (1) use of technology for radiation field reduction, (2) better dosimetry and accurate reporting of dose, (3) radiation protection planning, (4) exposure control during the execution of jobs. A general culture of supporting effective radiation dose management is also an important part of all of these methods.

4.1 Radiation Field Reduction

The most significant radionuclides that contribute to worker doses at nuclear power plants are the radionuclides 6oCo and “Co, which are responsible for over 80% of out-of-core radiation fields. Cobalt-60 has a 5.271 year half-life and produces penetrating gamma rays (1.173 and 1.332 MeV, both 100%) through its radioactive decay. Cobalt-60 is

produced by the bombardment of its precursor, 59Co, with thermal neutrons: 5yCo(n,y)60Co. Cobalt-59 is the naturally occurring 100% abundant isotope of cobalt. Cobalt-58 is produced from ‘*Ni by the fast neutron reaction 5*Ni(n,p)58Co. Nickel is a major constituent of the structural alloys used in the primary coolant system. Cobalt-58 has a short half-life (70.88 day) and its decay produces a 0.8 1 MeV gamma ray. It contributes substantially to PWR doses in early plant life.

Cobalt is present in nuclear reactors as a low-level impurity (~0.2%) in primary system construction materials and at high levels (-50%) in the hardfacing alloys used in applications requiring resistance to mechanical wear [Ocken, 19851. Cobalt is also found as an impurity in the Inconel-600 and 690 used as steam generator tubing, in the Inconel- 7 18 used as grid spacers in PWR fuel assemblies, in the Type 304 stainless steel tubing and sheathing used in BWR control blades and recirculation lines, and in the Type 304 stainless steel and Inconel -625 used in PWR control rod cladding. The cobalt-base alloys are used in applications requiring superior wear resistance, such as bearing surfaces, valve trim, and reactivity control mechanisms. The most popular hardfacing alloys, which contain 50 to 60wt% cobalt, are generally referred to as StelliteTM (a trademark of the Deloro Stellite Corporation).

Radionuclide precursors are released by corrosion or wear from components in contact with the primary coolant. They deposit on fuel rod surfaces in the reactor core, where they are activated. The resulting radionuclides are then transported by the coolant from the core, and deposit on out-of-core surfaces. They deposit either in the form of insoluble or colloidal particles or by incorporation into the corrosion film as an ionic form. Cobalt buildup is approximately proportional to corrosion rates [Wood, 19951.

4.1.1 Controlling Cobalt Sources

The largest contributors of cobalt sources were identified as the corrosion release of cobalt from the Inconel steam generator tubing and the wear and corrosion of control rod drive mechanisms and valves in PWRs [Young, et al., 1982; Bergmann, et al., 19821. In BWRs, 90% of the cobalt was attributed to the cobalt-base hardfacing alloys [Falk, 19821.


Controlling the cobalt source for steam generator tubing is not practical unless the entire steam generator is replaced. Indeed, a number of PWRs have replaced steam generators to address various forms of corrosion-induced degradation. The favored alternative to Inconel-600, the most common steam generator tube material used in the earliest generators, is Inconnel-690. This new alloy is found to exhibit much higher resistance to corrosion than Inconel-600. Typically, the residual cobalt content of Inconel 690 is about a factor of three lower than Inconel-600, which also serves to reduce the amount of cobalt available for subsequent activation [Ocken, 19971.

Brazed Inconel grids on the fuel assemblies of PWRs may contain up to 3.4 percent cobalt. Fuel assemblies need to be changed during normal operations to fuel assemblies containing low-cobalt grids, as appropriate. This provides an opportunity to use either Inconel with lower cobalt contents or Zircaloy grids, which have a very low cobalt content. The choice of a grid spacer alloy rests with the fuel vendor.

In an attempt to reduce wear and corrosion of cobalt-base hardfacing alloys as a radiation source, an iron-base hardfacing alloy has been developed as a generic substitute for Stellite. This is marketed as NOREMTM. NOREMTM alloys provide outstanding resistance to galling (adhesive) wear and cavitation erosion wear [Ocken, 19991. Economic considerations are not likely to justify the replacement purely for radiation control purposes. But for valves that are refurbished or replaced, seats, disks and other hardfaced parts, NOREMTM is recommended. An economic analysis [Welsh, 19961 showed that the potential for exposure reduction is significant only if the valve releases cobalt at high rates and personnel exposure at the unit is high. In addition, the cost of refurbishing large valves is disproportionally higher than refurbishing small valves. A review of all relevant data concluded that no limitations were placed on NOREMTM when it was selected for nuclear plant stop and swing check valves, globe valves, and butterfly valves [Ross, 19991. However, Stellite was still the preferred choice for valves characterized by large contact areas and long stroke length. Also, in tests of a 6-inch gate valve, with NOREMTM hardfacing, the leak rates were higher than a standard specified by Electricite de France (EdF) performed under simulated PWR operating conditions [Deicer, 19981. This indicates that caution is appropriate when specifying NOREMTM hardfacing alloy for use in some gate valve applications. A primary concern surrounds gate valves where design and service conditions can lead to large contact areas between two surfaces hardfaced with NOREM.

A guideline on the reduction of cobalt sources in nuclear plant valves [Ocken, 19931 provides a road map to deal with this complex issue. It indicates that a review of hardfacing attributes and the valve design and the duty to which the valve is subjected during operation are important considerations in selecting the optimum hardfacing alloy. In newer German PWRs, other cobalt-free alloys have been widely used as a hardfacing in nuclear plant valves. Subsequent dose rate measurements in these Konvoi plants show lower values than in earlier designs, confirming the value of reducing cobalt sources.

4.1.2 Preconditioning of Replacement Components

Preconditioning the surfaces of replacement components can reduce radiation-field buildup by rendering them less susceptible to corrosion and subsequent activity


deposition. Various surface preconditionings have been developed, including mechanical polishing, electropolishing, and pre-oxidation [Wood, 1995a].

Mechanical polishing serves to remove the oxide layer and is a good first step to achieve surface smoothness. Mechanical polishing is applied in several successive steps, using successively finer grades of abrasive grinding materials such as flapper paper or belt cloth with aluminum oxide. Electropolishing is accomplished by controlled anodic dissolution of the metallic surfaces, using an electrolyte and a cathode suitably shaped to accommodate the geometry of the component surface [Asay, 19901. Mechanically prepared surfaces without electropolishing are not as effective in reducing radioactivity buildup, and in some cases, may result in higher radioactive buildup regardless of the degrees of mechanical finish.

Pre-oxidation refers to prefilming surfaces to give a protective oxide layer before exposure to reactor water. This protective film effectively passivates the surface during subsequent exposure. Electropolishing has been widely used to treat the surface of replacement steam generator channel heads in PWRs and recirculation piping in BWRs. Electropolishing typically reduces activity deposition by a factor of two to three compared with a mechanically-polished surface.

Significant additional benefits can be provided by the use of advanced passivation technology. A very promising technique consists of electropolishing, followed by electroplating of a very thin layer of chromium, which is then followed by preoxidation in moist air [Wood, 1995a]. Dose rate measurements from replacement components treated with this process, including recirculation piping runs in BWRs and manway seal plates in PWRs, show reductions in activity buildup approaching a factor of ten lower than those measured on comparable electropolished surfaces [Galbraith, 19991.

4.1.3 Primary Chemistry Control

Occupational exposure in nuclear power plants is, to a great extent, due to the result of the buildup of radiation fields on ex-core surfaces caused by activated corrosion products. Minimizing corrosion product input and impurities in primary coolant water will minimize activation and deposition of radionuclides. Control of primary water chemistry has a significant impact on the release and transport of corrosion products. At the same time, the control of water chemistry affects the integrity of the materials and fuels in the system. Therefore, optimum water chemistry control must maintain a balance among fuel integrity, materials integrity, and reducing radiation fields [Wood, 1995a].

In PWRs, the pH and the lithium and boron concentrations are the parameters used to provide primary coolant chemistry control [Wood, 1995a]. Primary system pH greatly influences the release of corrosion products from structural alloys, crud buildup on fuels, and radiation-field buildup on out-of-core surfaces. In the early 1980s coordinated lithium/boron chemistry was the recommended chemistry control regime for the majority of plants in the U.S. [Solomon et al., 19841. This regime involves coordinating lithium hydroxide and boric acid concentrations to maintain a constant pH of approximately 6.9 for most of the fuel cycle.


Later studies [Driscoll, et al., 19911 on changes in corrosion product deposition as a function of pH showed that the nuclide deposition was highest at the lower end (6.5) of the studied pH range and decreased as pH increased. The pH range between 7.2 and 7.5 was found to beneficial. The benefit of using high pH (6.9 to 7.4 at 300°C) was also demonstrated by plant experience [Wood, 1990a; Wood, 1995b]. Elevated lithium concentration (2.2 ppm initially or higher) is required to maintain pH in the presence of high boric acid concentrations at the beginning of a cycle. This new regime requires initial boron concentrations well in excess of the 1200 ppm typically used for annual cycles. The expected benefit of this approach is the lower build up of crud, but a potential adverse effect is the increased corrosion of Zircaloy fuel rod cladding due to elevated lithium concentrations.

Most PWRs now operate using the modified coolant chemistry regime, rather than the elevated lithium concentration regime [Wood, 1995a]. Under the modified regime, the lithium concentration is held constant at 2.2 (+/- 0.1 Sppm) until pH 7.4 is reached. The boron concentration steadily decreases from the initial 1200ppm as the pH increases from 6.9. There are limited data available on which to base operation above pH 7.4. To maintain the pH at 7.4 with the continued decrease in boron concentration, lithium concentration is reduced. With this modified coolant chemistry regime, dose rates are lower and Zircaloy corrosion concerns are minimal.

Maintenance of a hydrogen overpressure to keep dissolved oxygen concentrations low is also found to be useful for radiation field reduction [NCRP, 19941. Low concentrations of dissolved oxygen prevent the buildup of corrosion products on fuel and transfer to out-of- core surfaces.

Zinc injection (including depleted Zn-64 to minimize radioactive Zn-65 formation) has also been demonstrated as a means of radiation field control for both PWRs and BWRs [Kohlmann, 1999; McElrath, 20001. More field data is available from BWRs [Giannelli, 19991, which show that zinc injection at a level of - 5 - 10 ppb reduces shutdown radiation fields by about a factor of two to three. The original motivation for injecting zinc in PWRs was to mitigate stress corrosion cracking of nickel-base structural alloys, which involved injecting zinc at - 30 - 40 ppb. Later dose rate measurements at one PWR plant show that zinc reduces shutdown radiation fields, even at a fairly low concentration of - 5 ppb. Although the earliest plant demonstrations suggested an increase in corrosion of the Zircaloy fuel rod cladding, subsequent analyses suggest the observed changes were not a consequence of zinc additions.

In BWRs, normal water chemistry is oxidizing in contrast to the reducing chemistry environment of the PWR primary systems. BWR coolant chemistry control focuses on converting the primary system from the normal oxidizing environment to a reducing environment by hydrogen injection, thus reducing the corrosion potential of the system. The hydrogen will consume a major fraction of radiolytically generated oxygen species. Introduction of hydrogen water chemistry (HWC) was effective in reducing the intergranular stress corrosion cracking of austenitic stainless steel structural alloys (e.g., recirculation piping system). However, plant experiences showed that HWC was responsible for the increase of both the 16N and MCo activity which are the primary source of radiation fields in the BWR systems [Wood, 1995a; Wood, 1999b].


Zinc injection in BWRs served to reduce the increases in radiation fields that followed the introduction of HWC [Giannelli, 19991. A recent work [Hettiarachchi, et al., 2000; Hettiarachchi, 19991 has shown that the needed hydrogen concentration can be lowered if a mixture of noble metals (Pt and Rh) are injected into the coolant. At the lead demonstration plants, dose rates measured at the outage following the introduction of the noble metals have decreased.

4.1.4 Reactor Shutdown and Startup Chemistry

While operating chemistry control has historically been the focus of the effects of chemistry control on radiation fields, the effects of shutdown and/or startup chemistry has become much more prominent during recent years. Industry experience has shown that plants that have not properly established primary water chemistry during shutdown conditions have, in some cases, observed elevated radiation fields [Millett, et al., 19991.

Radiation-field increases that occur during a reactor shutdown generally result from a release of radioactive corrosion products from in-core regions and subsequent deposition in low points or low-flow ex-core regions of the reactor coolant system. Although the release of radioactive corrosion products is an unavoidable consequence of the shutdown process, the negative effects of these releases can be diminished by maximizing the decomposition and solubilization of radioactive corrosion products through controlled burst [Ocken, 19991.

To cause a controlled burst of corrosion products in PWRs, hydrogen peroxide is added to the primary system resulting in the oxygenation of the system when the temperature is below 93°C (200°F). This solubilizes the corrosion products. The 58Co and @‘Co in the mixture are cleaned up through the letdown system. Even though the addition of hydrogen peroxide can result in localized high radiation areas in other areas of the primary coolant system, as a result of the migration of the corrosion products removed from the reactor vessel, out-of-core radiation levels can be successfully reduced resulting in a reduction in worker dose.

If the hydrogen peroxide were not added, the corrosion product burst would occur spontaneously when venting the system or lifting the reactor head. At this time, the letdown cleanup system would not be as effective, and much of the corrosion product would remain in the system. Early addition of borates during the shutdown and careful control of reactor coolant temperature have also been used to control the corrosion product burst. Thus, the optimum shutdown methodology should be determined for each plant’s conditions through the combination of options [Wood, 19931.

In BWRs, slow shutdown rate can reduce radiation levels [Giannelli, 19991. Operating the reactor water cleanup (RWCU) systems at maximum capacity during cool down and maintaining primary system recirculation flow at that time also helps reduce crud redistribution.

The primary focus in startup chemistry control in PWRs is to provide a minimum time that reactor coolant is maintained in an acid reducing environment, minimize the lithium


concentration in the decay heat removal system, and minimize the nickel present in the reactor coolant prior to startup [Millett, et al., 19991.

4.1.5 Chemical Decontamination

Approximately one-fourth of U.S. PWRs have elected to decontaminate their containment buildings to allow access either in street clothes or in clean modesty garments. These activities show [Kelly, 19991 that benefits of containment decontamination far outweigh adverse impacts. Those plants that have achieved clean containment status are very supportive of the concept and would strongly resist returning to a contaminated containment status.

Decontamination refers to the removal of radioactive-bearing deposits from the interior surfaces of piping and components. Decontamination has proven to be an efficient way of controlling radiation exposure [Wood, 1990b; Smee, 19991. Chemical decontamination uses chemical solvents to loosen and remove the corrosion product films. For nuclear plant system decontamination, relatively dilute reagent solutions that could be processed effectively by ion exchange resins for the removal of radioactive waste are used. Three dilute chemical reagents (i.e., CAN-DEREM, LOMI, and Citrox [Smee, 19991) have been used to perform the recent decontaminations at US light water reactors, while the CORED process is used on a regular basis in Europe. Decontamination technology is routinely performed, and provides an effective way for dealing with high shutdown radiation fields [Smee, 19991.

CAN-DEREM is a variant of the CAN-DECON process that does not contain oxalic acid. The CAN-DECON process is an acid dissolution process using an organic acid reagents based on citric and oxalic acids (e.g., CITROX), sometimes combined with a chelating agent such as EDTA to retain dissolved corrosion products in solution. The LOMI process is a reductive dissolution process using the low-oxidation-state metal ion which is a strong reducing agents.

LOMI is the process of choice for BWR recirculation piping, while the regenerable CAN- DECON process is favored for systems with high surface/volume ratios, such as heat exchangers. These processes provide decontamination factors of 5 to 15 [Wood, 1995a]. If high-chromium oxides are present, both types of process may require an oxidizing pretreatment. Alkaline and/or acid potassium permanganate is used for this purpose, dissolving chromium oxides by oxidizing them to soluble chromate.

Decontamination of a complete primary coolant system with the fuel removed was performed at Indian Point-2 in the US in March 1995. In this decontamination, approximately 4000Ci of activity were removed, using about 2000 cubic feet of ion exchange resin, and a decontamination factor of 7.8 was achieved. The utility estimates savings of over 8 person-Sv during the outage.

Economic evaluation suggests that full-system decontamination of a PWR with the fuel in place is cost effective. Visual examinations and nondestructive measurements of Zircaloy cladding oxide thickness showed that decontamination solvents do not have deleterious effect on fuel performance [Arlotti, et al., 19941. Efficient decontamination

42 M. 3. Yim and H. Ocken

does however require proper systems support such as the space, piping and hose connections, and ion-exchange resin columns.

4.1.6 Fuel Failure Control

Although fuel failures are not a significant contributor to radiation fields in general, plant operation with fuel defects increases primary piping radiation levels due to fission product buildup on piping [NCRP, 19941. In normal operations, fuel failures are limited to less than 0.1 percent.

With fuel failures, irradiated fuel fragments can escape into the work environment from failed fuel and then escape from the primary system (when open or otherwise breached for maintenance). These high-specific activity discrete radioactive particles (“hot particles”), although not a major source of occupational dose, result in high but localized skin doses. The occurrence of hot particle contamination can increase personnel exposure indirectly as a result of increased contamination controls and leads to less efficient operation and more complicated maintenance procedures.

Fuel defects can be minimized by careful control of reactor water chemistry and power level changes, and preventing any loose parts or foreign objects from getting into the primary system during maintenance and refueling operations.

4.1.7 Use of Fine (Sub-micron) Purification Filters

A substantial fraction of the particulates and debris in the primary system exists as sub- micron particles. Installation of sub-micron mesh filters during operation, upstream of letdown demineralizers and in purification of seal water for the reactor coolant pump in PWRs, has been shown to reduce in-reactor radioactivity and radiation levels in primary piping [NCRP, 19941. To control the burst of particulates during start-up and shutdown, absolute filters of 1 micron mesh can alternatively be used. The radiation levels on the filters have not been very high [Comley and Roofthooft, 19881. However, the buildup of 14C activity in these filters may require additional considerations in waste disposal [Miller, 20001.

4.2 Accurate Dose Reporting

As the basis of proper radiation dose management, accurate dose reporting is important. Accurate dose reporting not only requires the use of accurate dosimetry devices but correct use of these devices.

Experiences at nuclear power plants show that the use of film badges for worker personal dosimetry for gamma and beta radiation can significantly overestimate the dose [Renn, 19951. Comparison of dose results between the use of film badges and Siemens Electronic Personal Dosimeters (EPDs) at the Sizewell B PWR station in UK showed that typically the monthly collective dose as measured by film badges was a factor of twenty higher than the dose measured by EPD over the same period [Renn, 19951. EPDs also provide the real time display of dose and the alarming capabilities. For the neutron personal dosimetry, the track etch device using poly-ally1 diglycol carbonate (PADC) is


known to have advantages over other neutron personal dosimeters of sensitivity, energy dependence of response, insensitivity to photons and ability to withstand harsh environmental conditions [Tanner, et al., 19951.

It has been also shown that dosimeter placement affects the effective dose equivalent (EDE) measurement. For example, the widespread practice of supplementing a single front-worn dosimeter with an additional dosimeter placed facing a radiation source can significantly overestimate EDE [Owen, 19981. Researches using a Monte Carlo photon transport analysis and anthropomorphic phantom showed that EDE for external exposure varies depending upon the direction of the beam striking the body [Reece and Xu, 19971. According to these researches, beams striking the front of the body normal to the body’s major axis (i.e., straight on) produce the largest effective dose equivalent. The next highest EDE is produced by beams striking the rear of the torso. EDE falls significantly if the incident radiation departs from these two orientations. For point sources in contact with the body, the EDE is highest for males when the source is on the front of the torso near the gonads. For females the highest dose occurs when the point source is on the front of the torso near the sternum.

4.3 Health Physics Planning

Careful planning for radiation work is very important to achieve radiation dose reduction. The principles of time, distance and shielding must be properly implemented in the planning of outage work, along with the use of appropriate engineering controls based on the ALARA principle [NCRP, 19941. This requires the estimation and review of the expected collective dose with the definition of job scope. Tasks can be prioritized based on the results. Availability of radiological support is also important.

Mock-up testing and training can be very useful to familiarize workers with the work and procedures and to verify the operability of the special tooling. The training leads to reduced work time and worker dose. Robotics and automated equipment should be used where practical.

Robots can be useful for high dose jobs and may be cost-effective as a replacement for skilled labor in a tight labor market. The robots include automatic machines for refueling, remote handling of control rod drives (CRDs), reactor pressure-vessel stud tensioning and cleaning, semi-automatic CRD disassembly and cleaning, remote lapping of valves, remote controlled inspection and monitoring devices, and use of an underwater disassembly and cleaning device for control drives [NCRP, 19941. Robots have been applied in steam generator channel head entries, reactor coolant pipe welding, decontamination, underwater work and surveillance in high-radiation areas.

Proper portable and area radiation survey instrumentation must be made available to check for changes in radiological conditions during the execution of job. Neutron surveys must be performed in work areas where neutron radiation may be present. Availability of visual and/or audible alert/alarm systems with the use of electronic dosimeters at a preset dose or dose rate should be considered for high dose jobs. These devices increase worker awareness of their accumulated radiation dose and the local dose rate, and alert them to any unexpected changes in dose rates.


Protective clothing can be used in the radiation controlled areas to avoid skin contamination. At the same time, use of protective clothing can lead to decreased efficiency for the workers, resultant longer stay times and more dose, increased physiological stress and risk of industrial accidents or heat exhaustion, and radioactive waste. A study [Cardarelli and Sammon, 19971 showed that protective clothing causes stress on the cardiorespiratory system of a worker even during early exercise, when the subjects are well below heat exhaustion thresholds. Therefore, the use of protective clothing has to be carefully reviewed in comparison to other alternatives such as appropriate contamination control and/or engineering controls.

Full-face piece air purifying respirators can be an effective means of decreasing worker inhalation of airborne radioactive particles in the nuclear power plant environment if properly used. A study indicated that the use of respirators did not increase the time workers need to perform certain tasks (a bolt tightening task requiring 350 ft-lbs of torque) [Cardarelli and Woldstad, 199.51. Since the study was not designed to assess the effect of respirators on worker efficiency on tasks requiring communication between workers, poorly lit work spaces, or cramped working conditions, use of respirators should be reviewed on a case by case basis.

Temporary shielding in various forms (lead blankets and bricks, hollow plastic, aluminum, or steel shields) should be used as applicable to reduce general area dose rates and hot spots [NCRP, 19941. Various forms of lead shielding including lead blankets and bricks are available for gamma shielding. Hollow shields made of plastic, aluminum or steel can be moved in an unfilled condition and then filled with water on location for neutron shielding. Free standing scaffolds can be used to support lead shielding without putting weight on piping. Mobile shield racks can be moved to a desired location to provide a temporary shield wall with hanging blankets.

Other engineering controls include the use of tents, containment bags and glove boxes or use of local ventilation [NCRP, 19941. Use of tents and other containment devices can be effective in limiting the high contamination areas to a relatively small size. This can also help to eliminate airborne radioactivity areas and allow general access. Use of local ventilation can also reduce airborne activity, help to minimize the need for respirators and can aid in keeping workers cool to avoid heat stress. This ventilation can be supplied by flexible hose extension from a building ventilation system or by providing a separate portable ventilation system involving a fan and appropriate filters. Since the setup and removal of equipment involves external dose to installers, the use of these techniques should be evaluated to achieve a net benefit. The buildup of contamination within the confinement should be monitored.

Decontamination of components/work area can be effective for labor intensive tasks in high radiation fields. Removing hot spots near work areas and general access routes, which have relatively high-dose rates in comparison to the local background, can be particularly effective.

Jobs should be carefully coordinated with plant system configuration to maximize the shielding effect from filled components. For example, work in the vicinity of PWR steam


generators should be performed with the secondary side water full to increase the shielding effect. Operation of auxiliary components that may affect the worker dose must be optimized to minimize their dose during the work activities.

4.4 Exposure Control During the Execution of Jobs

Exposure control during the execution of jobs requires the application of the ALARA principle at the worker level. This requires the support and coordination of many functional groups including maintenance, operations, engineering and construction, and radiation protection.

A pre-job briefing covering the scope of work and the radiological conditions in the work area is important and requires the participation of actual personnel performing the work and the radiation protection personnel monitoring the job.

To assure that the job continues with the minimum practical exposure, the progress of ongoing work must be carefully monitored. The actual accumulation of dose against their estimated dose can be tracked by using proper monitors. The workers must be updated on current radiological conditions during the work. The work must be performed as quickly and accurately as possible to prevent having to repeat the job. Proper tools must be available at the job site. Dedicated tools can be specified for certain repetitive jobs. Special tooling can be developed for specific jobs to reduce time in high radiation areas and to increase distance from the radioactive source. If necessary, decisions to change the scope of a job must be carefully discussed with radiation protection personnel, and the necessary radiological precautions should be applied. The work group supervisor is responsible for reviewing plans, procedures, equipment and the work environment to identify techniques to reduce dose and the spread of radioactivity. Good communication and teamwork is important in this respect. Radio and video communication between the radiation protection technicians, workers, and the job supervisors can be effectively used to transmit work instructions and protection practices to the workers remotely and rapidly and to minimize entries to radiation areas.

As post-job activities, cumulative dose by job must be recorded and compared to the estimated dose or historical levels and to those experienced by other utilities to provide data for similar work in the future and to determine the effectiveness of the ALARA program. For the protection of transient subcontractor workers who move frequently around the country, a centralized data system can be used [Daubert, 19951. By centralizing the data from all of the units and regrouping all doses received by anybody on any of its sites, an accurate picture of every worker dose can be obtained.

Detailed post-job review can be performed for the personnel who participated in the work, the collective dose expended, radiological conditions throughout the course of the job, dose-reduction techniques used and their effectiveness, problem areas, and recommendations for improvements. The recommendations to improve dose reduction and limit the spread of radioactivity should be documented for future use. For repetitive jobs, a job history file should be maintained to document radiation protection techniques used, problems encountered and corrective actions taken.


A package of ALARA procedures should be maintained and should contain information in an easily retrievable format for review during future similar work. Reports on ALARA procedures implemented are useful to summarize and document information for management on the effectiveness and accomplishments of the ALARA program. A review of Radiation Work Permits and supporting surveys may be useful to determine the effectiveness of dose and contamination controls.

4.5 Implementation of a Radiation Protection Culture

Individual awareness of the importance of radiological protection is by far the most important basis for prudent radiation dose management [Boehler, 19951. This requires the knowledge and skills for radiological protection as provided by theoretical and practical training, through good distribution of information and by extensive experience feedback and the commitment of the management to a policy of radiological protection and the compliance of individuals with the aim of limiting exposure as far as is reasonably possible. Post job reviews/feedback, suggestions, audits are often used to assess the effectiveness of radiological controls and worker exposures. Motivation for achieving the best radiation dose reduction must be provided and reinforced through the personal responsibility and self-discipline of all those concerned with radiological protection. Proper supervision and monitoring and audits must be performed through a clear definition of the respective roles and responsibilities of those concerned from the operator to external contractors.

This requires a culture of radiological protection. This culture is as much a question of attitude as of structure, which concerns both individuals and organizations and which requires the on-going commitment of managers in the design and operation of nuclear installations. At the same time, the cost of achieving the required level of performance in radiation protection should be optimized. Many US utilities demonstrate the success of radiation dose management from their experiences. The dissemination of a radiological protection culture will also contribute to improving the company’s public and internal image.

5. Concluding Remarks

This study reviewed various approaches available for radiation dose management in nuclear power plants. These approaches include radiation field reduction, accurate dose reporting, good health physics planning, exposure control during job execution, and implementation of a radiation protection culture. To achieve radiation field reduction, the following measures can be implemented: (1) controlling cobalt sources through the use of low cobalt impurity materials in the structures and piping and iron-base hardfacing alloy in valves; (2) preconditioning of replacement components using mechanical polishing, electropolishing, advanced passivation technology, and presentation-oxidation; (3) controlling primary chemistry based on the adjustment of pH, lithium concentration and boron concentration in PWRs and the use of HWC with the injection of zinc and noble metals in BWRs; (4) controlling reactor shutdown and startup chemistry; (5) chemical decontamination using various chemical solvents; (6) minimizing fuel failures, and (7) using sub-micron purification filters.

Radiation dose management 4-l

US experience shows the factors contributing to savings in radiation exposures for the period of 1985 to 1990 are (in the order of rank): (1) development and implementation of radiation protection guidelines, (2) reductions in unscheduled special maintenance, (3) radiation field reduction (through material replacement), (4) cobalt substitution, decontamination, (5) chemistry control guidelines, and (6) automated in-service inspection methods, and heat stress management guide.

These examples illustrate that for successful radiation management various factors must be integrated if plant ALARA goals are to be realized. These factors are good planning and programming, use of available and proven technology, the culture to support ALARA among the management and workers, and carrying out the actual tasks according to specified details. Both the control of radiation fields through the implementation of technology and the control of human involvement in radiation zones based on application of the ALARA principle at the worker level are crucial to success.

48 M.-S. Km and H. Ocken

References

Arlotti G., Bauer F., Corpora G., Peffer D., Porterfield J., Pritchett P., Remark J., and Sejvar J. (1994) PWR Full Reactor Coolant System Decontamination Qualification Program: Decontamination and Reirradiation of PWR Fuel Assemblies, Electric Power Research Institute Report EPRI TR- 1038 16.

Asay R., Graves 0. , Guastafero C. T. and Spalaris C. N. (1990) Electropolishing Process Deveopment: PWR Channel Head, Electric Power Research Institute Report EPRI NP- 6619.

Baum J. W. and Matthews G. R. (1985) Compendium of Cost-Effectiveness Evaluation of Modifications for Dose Reduction at Nuclear Power Plants, Nuclear Regulatory Commission Report NUREG/CR-4373.

Bennett M. (1996) Transforming Dose Management Techniques through Technology. Nuclear Energy 35, 2, 105-109.

Bergmann C. A. and Bencini R. L. (1997) Evaluation of PWR Radiation Fields: 1991- 1996, Electric Power Research Institute Report EPRI TR-107566.

Bergmann C. A., Landerman E. I. , Lorentz D. R. and Whyte D. D. (1982) Evaluation of Cobalt Sources in Westinghouse-Designed Three- and Four-Loon Plants, Electric Power Research Institute Report EPRI NP-268 1.

Boehler M. (1995) From a Regulatory to a Cultural Approach in the Field of Radiological Protection, Radiation Dose Management in the Nuclear Industry, 13-l 9. Windermere, Cumbria, UK, British Nuclear Energy Society, 9-l 1 October.

Cardarelli R. and Sammon M. (1997) Effects of Protective Clothing on the Cardiorespiratorv System of Workers at Elevated Temperatures, Electric Power Research Institute Report EPRI TR- 1077 10.

Cardarelli R. and Woldstad J. (1995) Effects of Respirators on Worker Efficiency, Electric Power Research Institute Report EPRI TR-105350.

Comley G. C. W. and Roofthooft R. (1988) Recent Chemistry Studies at the Belgian PWRs, Doe1 3 and Doe1 4, Proceedings of EPRI Seminar on PWR Water Chemistry and Radiation Field Control, Palo Alto, CA.

Daubert G. (1995) An Efficient Tool to Follow Up Doses of Transient Workers in French NPPs: DOSINAT, Proceedings of the conference, Radiation Dose Management in the Nuclear Industrv, 99-103, Windermere, Cumbria, UK, British Nuclear Energy Society 9- 11 October.

Deicer F. (1998) Monitoring Performance of Valves with NOREM Hardfacin&Electric Power Research Institute Report EPRI TR- 109345.

Radiation dose managemenf 49

Dionne B. J. and Baum J. W. (1985) Occunational Dose Reduction and ALARA at Nuclear Power Plants: Study on Hiph-Dose Jobs, Radwaste Handling. and ALARA Incentives, US Nuclear Regulatory Commission Report NUREG/CR-4254, BNL- NUREG-5 1888.

Driscoll M. J., Harding 0. K., and Kohse G. E. (1991) In-Pile LOOP Studies of the Effect of PWR Coolant PH on Corrosion Product Radionuclide Deposition, Electric Power Research Institute Report EPRI TR- 100 156.

Falk C. F. (1982) BWR Cobalt Source Identification, Electric Power Research Institute Report NP-2263.

Galbraith G. (1999) Assessment of Chromium Coating Technology, Electric Power Research Institute Report EPRI TR-112982.

Giannelli J. F. (1999) BWR Activity Control - Plant Demonstration Results, Electric Power Research Institute Report EPRI TR-11298 1.

Hettiarachchi, S. (1999) Shutdown Dose Rate Behavior after NobleChemTM, Proceedings: Radiation Exposure Control Seminar, Electric Power Research Institute Report EPRI TR- 114003.

Hettiarachchi S., Cowan R. L., Law R. J., Miller W. D., and Diaz, T. P., (2000) NobleChemrM Application to BWRs for IGSCC Mitigation and Dose Rate Control, ICONE 8.

ICRP (International Commission on Radiation Protection) (199 1) Recommendations of the International Commission on Radiological Protection, Pergamon Press, Oxford.

Kelly J. (1999) Containment Decontamination Study: Review of Benefits and Impacts of PWR Containment Decontamination During Maior Outages, Electric Power Research Institute Report EPRI TR-112921.

Khan T. A. and Baum, J. W. (1988) Worldwide Activities on the Reduction of Occupational Exposure at Nuclear Power Plants, U.S. Nuclear Regulatory Commission NUREG/CR-5 158, BNL-NUREG-52086.

Kohlmann R. H. (1999) Experience with Depleted Zinc Oxide Iniection in BWRs: 1999 Progress Report, Electric Power Research Institute Report EPRI TR- 10874 1.

McElrath J. W. (2000) Zinc Addition at the Palisades PWR to Reduce Shutdown Dose Rates Electric Power Research Institute Report EPRI TR- 1000 190. -7

Miller C. C. (2000), Personal Communications, Pacific Gas and Electric Company

Millett, P. et al, (1999) PWR Primary Water Chemistry Guidelines, Revision 4, Volume 2, Electric Power Research Institute Report EPRI TR- 1057 14-V2R4.

50 M.-S. Km and H. Ocken

National Council on Radiation Protection and Measurements (NCRP) (1994) Dose Control at Nuclear Power Plants,” NCRP Report No. 120.

Ocken H. (1993) Cobalt Reduction Guidelines, Revision 1 Electric Power Research Institute Report EPRI TR-103296.

Ocken H. (1995) Reducing the Cobalt Inventory in Light Water Reactors, Nuclear Technoloa68, 18-28.

Ocken H. (1997) Radiation Field Control Manual, Electric Power Research Institute Report EPRI TR- 10799 1.

Ocken H. (1999) Performance of NOREM (TM) Hardfacinn Alloys, Electric Power Research Institute Report EPRI TR-112993.

Ocken H. and Wood C. J. (1992) Status Report on Radiation Exposure Reduction at US Nuclear Power Plants, Proceedings of the Water Chemistry of Nuclear Reactor Systems 6, 2, 1-8, British Nuclear Energy Society.

Owen D. (1998) Criteria and Methods for Estimating External Effective Dose Equivalent from Personnel Monitoring Results: EDE Implementation Guide, Electric Power Research Institute Report EPRI TR- 109446.

Phillips M. K. and Findlan S. J. (1992) Welding of NOREM Iron-Base Hardfacing Alloy Wire Products: Procedure for Gas Tungsten Arc Welding, Electric Power Research Institute Report EPRI TR-101094.

Reece W. D. and Xu X. G. (1997) Determining Effective Dose Equivalent for External Photon Radiation: Assessing Effective Dose Equivalent from Personal Dosimeter Readings, Radiation Protection Dosimetrv, 69, 3, 167-178.

Renn G. (1995) Sizewell B Power Station Control Dosimetry System, Radiation Dose Management in the Nuclear Industry, 104-106, Windermere, Cumbria, UK, British Nuclear Energy Society, 9-l 1 October.

Robinson I. F. and Lazo E. (1995) ISOE: An International Occupational Exposure database and Communications Network for Dose Optimization, Radiation Dose Management in the Nuclear Industry 92-98, Windermere, Cumbria, UK, British Nuclear Energy Society, 9-11 October.

Ross A. M. (1999) Compilation and Evaluation of NOREM (TM) Test Results: Implications for Valve Applications, Electric Power Research Institute Report EPRI TR- 109343.

Smee J. (1999) Decontamination Handbook, Electric Power Research Institute Report EPRI TR- 112352.


Solomon, Y., Shaw R. A., Roesmer J. , and Bergman C. A. (1984) Crud Effects and PWR Coolant Chemistry: State of the Art, Water Chemistrv of Nuclear Reactor Systems 3, 2, 35-42, British Nuclear Energy Society.

Tanner R. H., Bartlett D. T., Gilvin P. J. and Lavelle J. (1995) The Use of PADC for Neutron Personal Dosimetry in the Nuclear Industry, Radiation Dose Management in the Nuclear Industry, 79-85, Windermere, Cumbria, UK, British Nuclear Energy Society, 9- 11 Octorber.

Thezee C., Sticker L., and Do110 R. (1995) The EDF ALARA Projects: A Global Management Approach, Radiation Dose Management in the Nuclear Industry, 28-32, Windermere, Cumbria, UK, British Nuclear Energy Society, 9-l 1 Octorber.

Welsh J. (1996) An Economic Analvsis of Cobalt Valve Replacement, Electric Power Research Institute Report EPRI TR- 107 169.

Wilson R., Vivian G. A., Chase W. J., Armitage, G., and Sennema, L. J. (1986) Occupational Dose Reduction Experience in Ontario Hydro Nuclear Power Stations Nuclear Technology 72, 23 l-245.

Wood C. J. (1987) Radiation Field Control at LWRs: The End of the Beginning Nuclear Engineering International, 27-30, November.

Wood C. J. (1990a) PWR Water Chemistry Guidelines, Revision 2 Electric Power Research Institute Report EPRI NP-7077.

Wood C. J. (1990b) A Review of the Application of Chemical Decontamination Technology in USA Progress in Nuclear Energy, 23,35-80.

Wood C. J. (1993) PWR Primary Shutdown and Startup Chemistry Electric Power Research Institute Report EPRI TR-101884.

Wood C. J. (1995a) Dose Reduction Methods: Chemical and Physical in Radiation Protection at Nuclear Reactors, by C. J. Maletskos (ed.), Medical Physics Publishing, WI,

Wood C. J. (1995b) PWR Water Chemistry Guidelines, Revision 3 Electric Power Research Institute Report TR-105714.

Wood C. J. (1999a) Cost Effective Radiation Exposure in a Competitive Environment, Proceedings: Radiation Exposure Control Seminar, 1999 Electric Power Research Institute Report EPRI TR- 114003.

Wood C. J. (1999b) BWR Water Chemistry Guidelines 1999 Revision Proceedings: Radiation Exposure Control Seminar, 7-l-7-10, Electric Power Research Institute Report TR-114003.

Young T. R., LaDuca J. T., Quattrochi T. E. and Wiatrowski M. (1982) Cobalt Source Identification Program, Electric Power Research Institute Report EPRI NP-2685.

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Radiation dose management in nuclear power plants