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INTRODUCTION

A common practice in the nuclear industry involves locating radioactivity, and doing something with it, e.g., removing it. This might include packaging and disposition of radiological waste, decontamination of areas with radiological spills, maintenance operations, routine radiological assessments of high dose areas, and emergency response to unplanned events. The location and assessment of the radioactivity must be done under the proper conditions of worker safety and also done in the most economical and timely manner.

There are three tasks that must be done in the characterization and work-planning phase of each job. First the radioactivity must be located with adequate precision for the task. Next the nature of the radioactivity - nuclide and amount - must be determined. Finally, dose estimations must be developed in order to establish a work plan to do the job under appropriate levels of safety, cost, and schedule. For large projects, these subtasks of location, quantification, and dose estimation are repeated periodically throughout the project, and the data reviewed to see how it changes the work plan.

The purpose of the article is to present some new tools that can help in the radiological characterization and dose planning phase of the project to achieve the goals of economy, quality, and safety.

TRADITIONAL METHODS

Of course, for properly performed projects today, these tasks of location, quantification, and dose estimation are being done. But the tools and techniques to do them haven't changed much in the past 20-30 years.

Radioactivity location: There are two common ways. First is just to assume (guess) its location and distribution. This guessing will certainly be guided by detailed plant knowledge of where the radioactivity is supposed to be, and past experience from similar jobs of where it was then, and perhaps by some limited survey meter readings. Because we filled a drum or box with radioactive materials, it is common to assume that the radioactivity is uniformly distributed throughout the entire container. Because a certain pipe was used to transport radioactive fluids, it is common to assume that the distribution within the empty pipe is uniform. If there is a large grouping of items containing a certain piece of equipment that is supposed to have radioactivity in it, it is common practice to assume that dose-rate coming from the area is mostly from that piece of equipment. This informed guessing may indeed be correct, but often it is wrong, with the consequence of unnecessary cost.

The second method is to make detailed sets of measurements with field technicians using hand-held survey meters. For the situation of single well-isolated sources, and with well qualified and trained operators, the locations can indeed be found rather precisely. But, if there are multiple sources close together, this doesn't work very well as the probes used are typically designed to be anisotropic, and respond to radiation from all directions. Presenting this data to others is also normally necessary. Dose-rate computations need to be made for dose planning. Workers need to be trained before doing the job, to minimize dose. This requires lots of readings to be recorded, along with the precise location of the probe and the object. This recording can be very time consuming; lots of readings or dose-rate and probe coordinates. And then this data must be transcribed and mapped so that it can be interpreted by the planners and by the workers during their training. All of this takes time, which is expensive. And, much of this time is spent by measurement technicians in a radiological area, incurring avoidable dose. Finally, most of these measurements are made under less than optimum conditions of worker safety. Workers are clothed in special anticontamination clothing (gloves, shoe covers) that restrict free motion, commonly wear devices that restrict vision and breathing (face masks and respirators), and then told to go into a high temperature area and climb a ladder carrying a heavy instrument to take and write down lots of instrument readings. While the radiological safety is normally well controlled, there are improvements that can be made in general occupational safety.

Nuclide identity and quantity: Again, for most jobs the most common practice is assuming (guessing, again) that the nuclide is known, and then using simplistic doserate-to-activity conversion methods. This assumption is made either based upon the nuclides expected to be there from knowledge of how things were supposed to work, or better, from samples taken from previous jobs and laboratory gamma spectroscopy analysis.

Where conditions are so unknown that valid assumptions cannot be defended, then a detailed set of samples from this job are typically taken and analyzed in the laboratory. This sampling and laboratory analysis method also has significant time, dose, safety and quality problems. It takes time to extract a sample, package and transport it to the laboratory, prepare and analyze it, and then report the results. This elapsed time is very costly to most projects, and the labor to do it is also expensive. The worker that must extract the sample also incurs some dose. And, if the sample isn't from a location designed to be sampled, then extracting it can be quite difficult and dangerous. Drilling for cores of concrete, taking samples of steel, getting material out of the insides of pipes or tanks, or waste containers, etc., all have occupational safety problems. And doing this with radiological protective clothing, and on scaffolds or ladders just multiplies the risk. So as a consequence, to minimize the time, cost, and safety risks, not very many samples are taken, which impacts quality. For a nonhomogeneous source, which most are, many small individual samples must be taken to get a good statistically reliable value for the total activity. Of course, this could be improved by simply taking a few very large samples back to the laboratory, or even taking the entire item back to the laboratory.

Dose planning: The object is to minimize the dose to workers by proper job planning. To do this planning requires knowledge of dose-rates at areas where the workers are expected to be. If this is a radiologically static job where the dose-rates don't change during the job, this can be quite simple - just send a person in there with a survey meter to measure it. But, it isn't that simple for most dosimetrically important jobs. The dose-rate will change during the job, as a result of what the workers are doing. Temporary shielding is frequently added to reduce the dose-rate. Radioactive contamination or components are removed, changing the dose-rate. The ALARA concept requires evaluating these things and taking the appropriate action to keep the dose "reasonable". Dose planning is this evaluation.

Dose planning is done using the best available information about where the radioactivity is, what it is, and how much is there. Then, estimates are made as to how these parameters will change during the job. Finally, this is combined with occupation times for workers doing the job to estimate the individual and collective dose for that job. Various plans are evaluated to determine the optimum one. Obviously, the better this input information is, the more accurate the estimated dose-rate and planned dose will be.

Computer software or manual calculations are used to estimate dose-rates when measurements are not practical. This can be simple distance conversions, or more complex shielding calculations. But when there are multiple sources causing the dose-rate, or when the object causing the dose-rate is not a simple shape, these calculations and common software programs do not work well. And, creating a model using some of the more complex computer codes, can be rather daunting and time consuming, and still needs good input data. So it is not frequently done.

NEW TECHNIQUES

What would be quite helpful would be an instrument that tells the user where the radioactivity is, does it quickly, without dose or risk of injury, and presents the information in an easy-to-interpret format. It would also be useful to be able to determine the exact nuclides and the activity in each of the contaminated areas of the item or area, again quickly, at reduced dose and safety risk, and with improved quality. Finally, it would be nice if all of this information could be entered into a computer model that is capable of complex simulations but easy to use, to determine accurate dose-rates at many different locations, to allow simulations of various dose reduction techniques.

These techniques do indeed exist, and are commercially available from several established vendors for the past several years. Imaging systems that capture data and quickly present the user a radiological dose-rate map superimposed upon a visual image of the area are available from at least three different vendors (BNFL, EDO, and CANBERRA). Portable in situ gamma spectroscopy systems that can identify and quantify gamma emitters quickly and accurately, are available from at least two different vendors (Ortec, CANBERRA). Dose-rate calculation software is available from several commercial sources (Framatome-ANP, CANBERRA, probably others), as well as from codes from government sources like MCNP and QAD from US DOE. So this discussion is not about what techniques could be developed and implemented in the future, it is about what can be implemented now.

GAMMA IMAGING SYSTEM

All three commercially available instruments operate in a different technical manner, but all three instruments ultimately end up with a combined visual image of the area that was measured with an iso-countrate profile superimposed on it showing where the radiation came from. None of the instruments are very complicated to operate, but they are all more complicated than the survey meter they can replace, and considerably more expensive. The Cartogam is CANBERRA's offering as shown in Figure 1. It uses technology developed jointly by the French CEA, Cogema and CANBERRA Eurisys and adapted for commercial production by CANBERRA Eurisys.

The Cartogam detector unit is fundamentally a gamma sensitive pin-hole camera. A reversed image of gamma emissions from the detector's field of view is projected upon a gamma sensitive scintillator. This is exactly like the original "camera obscura" which was a precursor to today's photographic cameras, or like used to project images of a solar eclipse onto a screen for safe viewing. The gamma scintillator converts these photons into visible light. The visible light is amplified with an image intensifier and projected onto a CCD where it is recorded.

The Cartogam is physically small enough and light enough to allow it to be deployed by one person in a radiological area. The detector head unit is 8 cm diameter and 40 cm long, and weighs only 17 kg. It is connected to the control unit (a portable industrial PC) by a single cable that can be up to 250 meters in length. The only radiation exposure to the workers is for the technician to take the detector in, and aim it at the area to be measured. He then leaves and all further work is done at the remote PC. For multiple measurements, the detector head can be set up on a motorized pan/tilt unit with remote control. Figure 2 shows such an operation. Now the remote operator can collect a large number of measurements, under different instrument conditions and of different portions of the room, thus building an album of images that can be assembled into a mosaic of the entire room. Typically, a second set of measurements would be made from a different location to see the source from a different angle to allow its precise location in all three axes (including depth) to be determined.

Since most areas likely to be imaged also have loose contamination, an important characteristic of the Cartogam is its easy ability to be protected against contamination. The detector head is a simple smooth sealed cylinder. It is very easy to wrap with plastic to protect it, as there are no moving parts like cooling fans that cannot be blocked. The unit is sealed, and can be pressurized with an inert gas, or it can have a gas flow through it, to further minimize any in-leakage of contamination. The pan/tilt unit is also very easy to wrap for protection, as it only moves infrequently, not continuously like other raster-scan imaging devices.

Data analysis and basic interpretation is rather simple - just look at the image and see where the concentrated sources of emitted photons are. Figure 3 shows the type of output that is available. A single image can show the polar direction of the source from the detector, but cannot show the distance from the detector. And if there are two sources on the same radial angle, only one will be shown. Unless the distance to the source is obvious, a second measurement from a different angle would normally be made.

The output of imaging systems can be used directly, and also as input to other applications like in situ spectroscopy and dose modeling.

The benefits of imaging as compared to manually measuring the dose-rate and manually recording the data onto a graphic presentation include the following:

Lower dose

• Less time is spent in the area; fewer man-hours to gather the data
• The time spent is at a farther distance and in a lower dose-rate area
• There is less time spent performing non-measurement tasks
  • Less time to erect and remove work platforms or scaffolds
  • No time to record the data and the location
• Lower dose to workers after using the graphical images to understand where the sources are, and how to work better around them

Better industrial safety

• Less time scrambling about on scaffolding or ladders in clothing that is difficult to work in, and with devices that restrict vision, and/or in areas of elevated temperature
• Results available nearly immediately
  • No lost time transcribing the data onto a map of the area, if one exists
  • Within a few seconds of completion of the measurement, the user is presented with the graphical image of the area and the radiation emission locations

Better quality data

• One of the applications of the imaging data is worker training; the video image of the item better identifies it than the normal sketch; the superimposed iso-countrate contours are more easily interpreted than the normal dose-rate values written on a sketch

Of course, we recognize that imaging systems are not suitable for all applications, and cannot totally replace all traditional dose-rate measurements in any one application. Some benchmark measurements must be made, in order to add to and validate the conclusions that will be inferred from the imaging and in situ and dosemodeling tasks. But it can replace most of the dose-rate measurements in most of the applications where the dose to the people doing the measurements is a potential problem. And, since this dose-rate reduction is relatively easy to do, it should be done in accordance with the ALARA principle.

IN SITU GAMMA SPECTROSCOPY

Whereas gamma imaging systems are relatively new, in situ gamma spectroscopy has been around since the early '70s. The most common application was for environmental monitoring of soils, using the detectoron- a-tripod method, using the infinite field calibration method. But in the past five years, new advances in electronics and computers and analysis methods have made this much more practical for routine measurements inside nuclear facilities. These developments include:

Detectors

• High Purity Ge detectors (HPGe) have replaced the Ge(Li) detectors that could never be warmed up
• Although liquid nitrogen (LN) is required, it is commonly available throughout the world as a welding supply
• Detectors come in small packages that can be carried with one hand, but that hold five days of LN
• Multi-attitude LN containers that allow the detector to be pointed in all directions
• New type of Ge detector (the CANBERRA BEGe) that is good for almost all nuclides and applications - good efficiency and resolution at both low energy and high energy

Detector deployment systems

• Shielding systems that allow the detector directional response to be modified to a preferential direction by minimizing photons from other directions
• Detector holding mechanisms that allow the detector- shield system to be pointed in all directions
• Wheeled carts that allow the detector and shields to be relatively easily moved to the measurement location

Detector acquisition electronics

• Small portable battery operated Multi-Channel Analyzers (MCA) that can last all work-day on a single battery
• Digital Signal Processing electronics for excellent temperature stability and excellent performance at both low and high count-rates
• The CANBERRA InSpector™ 2000 is a good example of this technology
• Small portable highly powerful portable batteryoperated PCs that control the MCA and analyze the data from it

Spectral analysis software

• Gamma spectral analysis software to control the MCA, and to analyze the spectrum and generate nuclide identity and nuclide activity, e.g., CANBERRA's Genie™ 2000 software
• Mathematical efficiency calibration software which generates the conversion factors between photons measured and the activity that caused it, for a wide variety of object sizes and shapes; e.g., CANBERRA's ISOCS™ software

It is this combination of items that finally materialized in '97 that allowed the introduction of the first in situ system for in-plant applications, the CANBERRA ISOCS system. It has been quite well accepted, with hundreds sold and in use at operational nuclear plants and facilities undergoing remediation. Figures 4 and Figure 5 show the ISOCS shield system, the detector, the MCA, and depict the mathematical calibration.

Some projects proceed with just dose-rate measurements made near an item (pipe, valve, etc.) and then "plant knowledge" is used to guess at the true location of the radioactivity and the nuclides causing the dose. Then dose-to-activity conversion factors are used to estimate the activity causing the measured dose-rate.

This practice is flawed in three ways. First, the location is assumed rather than well known, but the use of imaging systems can help solve that problem. Second, the nuclides causing the dose are assumed rather than well known. Third, the activity is not well known due to the first two problems which are inputs into the dose-activity formulas. A good solution is to take samples to determine the nuclides and the activity. An even better solution is to use in situ gamma spectroscopy to determine the nuclides and activity. In addition to improving the knowledge of the factors causing dose in the project, in situ gamma spectroscopy can also save time and dose in the process. Figure 6 shows a typical pipe monitoring application.

Whereas the imaging system gives output showing where the radioactivity is, the in situ system gives output telling what nuclide is there, and how much is there. This new technique can replace or largely supplement the process of taking samples and analyzing them in the laboratory, or guessing at the nuclide and using dose-rate to activity conversion factors.

In situ systems have similar advantages over sampling as does imaging over manual measurements:

Lower dose

• Less time to set up and take measurements than to take samples
• The measurement task is done at a far distance, and thus in a lower dose-rate area than taking samples or dose-rate measurements

Better industrial safety

• No need to perform the somewhat dangerous tasks of sampling radioactive items, drilling in concrete, working with fluids that might be toxic or corrosive, taking samples of high pressure gases, etc.
• Since in situ measurements are generally made at a distance, there is reduced need for climbing on ladders or scaffolds to get to the measurement point

Results available nearly immediately

• Whereas sampling takes time, and then must be packaged and transported back to the laboratory for preparation and analysis, in situ spectroscopy results are available nearly immediately

Better quality results

• More accurate and more defendable than just assuming the nuclide and then using dose-activity formulae
• Most sources are somewhat non-homogeneous, consequently it is very difficult to extract a "representative" sample. To reduce this sampling error, either more samples can be taken or larger samples can be taken, both of which can cause more time, dose, and risk of injury. In situ spectroscopy essentially follows the "large sample" - assaying the entire item as a large sample

MODELING AND DOSE-RATE CALCULATION SOFTWARE

The most common computation methods fit into two categories. For modeling of complex items and structures, and to generate quite accurate results, software like MCNP and QAD (US DOE) or Mercure (French CEA) are used. The most accurate are Monte-Carlo programs like MCNP where all radiation interactions are well modeled and tracked. When properly applied, these programs are acknowledged to provide the most accurate results. But, these computations take lots of computer time on powerful (expensive) computers, especially when there are multiple sources, large sources, and well-shielded sources.

To speed computations up, programs like QAD and Mercure use a point-kernel method, which basically breaks the source into many different small point-like sources and uses attenuation calculations to estimate the dose to the receptor from each of these attenuated points. While this greatly speeds up the analysis, simple attenuation formulae do not account for scatter; an additional "build-up factor" must be applied as a multiplier to the attenuated dose. This multiplier is frequently in the 10-100 range, especially when there is significant attenuation of the source. So errors in determining this factor are the major item limiting accuracy of these point-kernel methods. The determination of the appropriate build-up factor to use can be a complex process, as it depends upon the energy, the thickness, and the type of attenuator. Multiple attenuators in series can be especially challenging.

But the major limitation to these programs is the complex data input methods. These programs were developed years ago, before Graphical User Interfaces were common. To describe an exposure geometry can be quite daunting for users who don't do shielding calculations as a full time job. One of the most popular solutions has been MicroShield, by Framatome-ANP. This code uses a series of pre-defined source shapes (points, lines, cylinders, boxes, etc.). The user then adds attenuators (as long as they fit their predefined shapes) and receptor locations for the dose-rate to be calculated. The software is quite simple to use, as long as the case fits the pre-defined models. The graphical user interface provides good feedback to the user about the correctness of the dimensional inputs. Limitations include scenarios where there are more than one source, and buildup factor accuracy.

Mercurad is a newly introduced product, that fits in the space between the popular but simplistic MicroShield and other accurate but user-unfriendly codes. Although the product is new, it is not a new computational code. The core of the computational code is the standard unmodified Mercure Version 6 software. It does all of the computations. What is new is the graphical user interface to allow the user to create complex objects. The user is not limited to predefined source shapes and pre-defined attenuator shapes. Figure 7 and Figure 8 show examples. The GUI also allows the user to rotate and move the scene or the observer, and also choose various types of visibility options to see all parts of the scene from all angles. And the user can add as many sources as appropriate to define the scene, as shown in Figure 8. Each source acts as an attenuator for the other sources. The other major improvement over other pointkernel codes is the automatic build-up factor. No longer does the user have to guess at it, or choose among the various attenuators for a compromise. An automatic algorithm is built-in which has been demonstrated to work even with multi-layer attenuators.

With Mercurad, the user can create complex scenes with multiple sources shaped in a realistic way. The input to create the physical shape of the items would come from engineering drawings and/or physical dimensional measurements. The location of the radioactivity comes from the imaging measurements, supplemented with survey meter countrate-vs.-position measurements. The nuclides and activity values come from the in situ gamma measurements, and/or sample measurement results. First computations should be to benchmark the model; see if the dose-rate measurements in key locations can be reproduced by calculations. If so, then the model is validated. The validated model can then be used for dose planning. Try adding various types and thickness' of shielding configurations to see what doserate reduction it provides in the areas which will be occupied by the workers. A large number of dose-rate points can be computed in a single run. Combine these results with occupancy times for the various tasks that are planned to estimate the man-Sv for the job. Try various ways and schedules to do the job to see which one is really ALARA. Now, this type of planning is easier than before.

EXAMPLES SHOWING SYNERGISTIC COMBINATIONS OF THESE NEW TOOLS

Shielding Calculations:

It is typical for many pipes and valves and equipment to be located in a small compact area, like an equipment galley. Periodically some kind of maintenance must be done in those, such as changing out a pump. To lower the dose-rate, temporary shielding is added. The areas to be shielded, and the amount of shielding are usually just estimated, based upon past experiences and expectations. Sometimes too little shielding is added, and more must be applied. Or, sometimes too much is added; too much area is covered, or too much thickness is added. All of these scenarios cause extra dose, and extra time to do the job.

A better method would be to first map out the area with a combination of imaging measurements supplemented by a few dose-rate measurements. The small compact size of Cartogam will be quite important here, as it can be easily and quickly moved into position, and then remotely moved for multiple measurements. Next, use in situ Ge spectroscopy to determine the nuclide contents and level of activity of each of the sources in the room. The imaging information will be used to plan the in situ measurement strategy. Quantify both the small area hot sources and the large area sources to attempt to account for most of the activity that causes dose. The ability of ISOCS to give accurate results with highly collimated Ge detectors will be important here as the dose-rate in these small rooms will be too high for unshielded or NaI detectors.

Figure 9 shows a customer's example where the item in the center was being measured with both ISOCS and Cartogam.

Finally, create the input model for the dose calculation software, using the imaging data, the in situ data, and the physical parameters of the items. Test out the model by computing dose-rate at the measured locations. The advanced GUI and ability to create complex geometries with Mercurad will be important here.

Next, use the model to try out various sizes, types, and thickness' of shielding. Make the shielding big enough to cover the sources, but not much bigger. Make it thick enough to provide the optimum level of dose-rate reduction; not too thick or not too thin as both will cause unnecessary dose. Use the imaging data as part of the training given to the installation workers to visually present to them the location of the hot sources; most workers absorb this highly realistic visual information much better than an engineering drawing with some dose-rate numbers written on it. After the shielding has been installed, do a limited set of imaging to confirm it was done properly and to see if other lower level sources are now visible, and if they are worth shielding or avoiding. Use these images as part of the training for the workers doing the pump changeout task.

Minimizing the cost to dispose of a large radioactive item:

Normally these items are not homogeneously radioactive; there are hot areas and other lower level or clean areas. First, use Cartogam to find out where the radioactive areas on the item are, and relative activities for each source. Then, use collimated ISOCS to find out what nuclides and how much for each of the interesting areas of radioactivity. Build a sufficiently detailed model of the item, using Mercurad. Validate the model by computation and comparison to measured dose-rates at a few locations.

Design the optimum shielding - less in some areas, more in others. Then compute the transportation and disposal costs. In addition, estimate the cost and dose to remove the hot portions of the item for separate shipment in standard containers and to minimize the portion of the disposal costs related to high activity or dose-rate. If portions of the item are low level or not radioactive, then fractionation of the large item might further reduce the shielding, transportation, and disposal costs. Where work is done, either removing items, or applying shielding, use Cartogam to train workers, and to evaluate shielding application. Use ISOCS to assay activity of removed items for waste disposal, and to help determine the required shipping container.

Figure 10 shows Cartogam imaging a waste container, which is then quantified by ISOCS.

Removal of concrete wall with embedded contaminated piping:

Use Cartogam to scan the wall to find out where the buried pipes are, and relative activities for each source. Use collimated ISOCS to find out what nuclides and how much for each of the interesting areas of radioactivity. Multiple measurements with ISOCS at various angles and the evaluation of nuclides at various energies can be used to give reliable estimates of the depth of the radioactivity in the pipe.

Build a model using Mercurad. Use Mercurad to evaluate the dose to operators with various decontamination or removal strategies. During the operation, use periodic Cartogam/ISOCS measurements to guide the work. The closer the contamination is to the surface, the more reliable the measurements.

Emergency response to a real or suspected "Dirty Bomb":

If there is suspicion of hazardous elements (i.e., explosives in this case, or chemicals or fire in the case of a transportation accident) then it is not prudent to send survey technicians up close for detailed measurements with conventional hand-held instrumentation. If there is some indication of radioactivity (e.g., increase in survey meter countrate as you approach the area) then use Cartogam to find out where the radioactive items are, and where they are not. Use ISOCS to determine what nuclide and what quantity. The use of Mercurad in the accident response phase is probably not likely, but its use in the planning phase is certainly appropriate. It is helpful to work through various emergency scenarios to estimate doses from example recovery actions, and to preplan various shielding options. Use the Cartogam image to brief the bomb disposal person (or fireman) on where he can most safely be, and then tell him to worry about the bomb, and not the radiation. The bomb will kill him immediately if he makes a mistake, but the radiation won't or at least it will take many years.

Sorting and disposal of legacy waste:

Here, there are many containers of radioactive waste that were packaged many years ago, without adequate data about the contents. External dose-rate measurements on the containers are quite variable, so it is likely that the actual items causing the high dose-rates are small in size, and can be re-packaged for more optimum disposal. The facility will build a special sorting area where the containers will be opened and the contents spread on a sorting table. A Cartogam will view the sorting table and identify the most radioactive items. Remote manipulators will remove them and then the area will be imaged again to see if anything was hidden. An ISOCS will be used to identify and quantify the high dose-rate items, individually to determine their ultimate disposition. The low level items are consolidated into a waste shipping container, and then assayed in the container. Mercurad will be used to create any special shielding for interim storage and shipping of the highly radioactive items.

CONCLUSION

There is a continual evolution of tools and techniques in any business. The business of operation and decommissioning of nuclear facilities is no different. Methods that are shown to be beneficial will ultimately supplant less effective techniques. The three new techniques presented here (gamma imaging, in situ gamma spectroscopy, computer modeling and dose-rate computation) are all relatively new in the nuclear maintenance business. And they are all technically defensible in the results that they generate; they can all be used today whenever the licensee thinks that they are beneficial. These new tools will never completely replace the conventional methods of using field technicians to measure dose-rate, and using sampling and laboratory instrumentation to determine nuclide identity and quantity, and using past experience to do dose planning. But, as these tools are tried, and as users become more qualified and experienced in using them, and as these tools and techniques improve with successive versions, they will change the way things are being done, and they will reduce worker dose, get jobs done sooner, reduce worker accidents, and provide better documentation of the work that was done.